Journal of Nuclear Materials最新文献

英文中文

Fuel performance simulations of TRISO particle geometries derived from XCT 从 XCT 得出的 TRISO 粒子几何形状的燃料性能模拟

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

Journal of Nuclear Materials

Pub Date : 2025-02-26 DOI: 10.1016/j.jnucmat.2025.155714

M. Poschmann, A. Prudil, R. Osmond

The current AGR TRISO fuel specification effectively assumes that the layer thickness variations within a particle do not significantly affect particle performance. However, the limits of this assumption and their relevance for commercial TRISO production have not been established. In this work, a method was developed to generate 3D geometries of TRISO particles, including the spatial variation in layer thickness, from X-ray computed tomography for use in fuel performance modelling. Simulated irradiation of a demonstration particle found SiC hoop stress values peaking at 315 MPa in tension, significantly in excess of those from previous modelling studies with similar particle aspect ratios. Simulations with representative 2D axisymmetric geometries based on the demonstration particle predicted significantly lower stresses for the same simulated irradiation. 2D radial segments extracted with an arbitrarily oriented polar axis under-predicted the maximum SiC hoop stress by 315-400 MPa, while those extracted with the polar axis passing through the point of maximum SiC hoop stress in the 3D model under-predicted the maximum SiC hoop stress by 165-275 MPa. The 2D model produced using existing methods for generating a 2D flat-spot particle under-predicted the maximum SiC hoop stress by 215 MPa. These findings suggest that existing models may underestimate the stress caused by the asphericity of certain TRISO particle morphologies, and that the current AGR specification may not capture all of the geometric factors that contribute to particle failure probability.

{"title":"Fuel performance simulations of TRISO particle geometries derived from XCT","authors":"M. Poschmann, A. Prudil, R. Osmond","doi":"10.1016/j.jnucmat.2025.155714","DOIUrl":"10.1016/j.jnucmat.2025.155714","url":null,"abstract":"<div><div>The current AGR TRISO fuel specification effectively assumes that the layer thickness variations within a particle do not significantly affect particle performance. However, the limits of this assumption and their relevance for commercial TRISO production have not been established. In this work, a method was developed to generate 3D geometries of TRISO particles, including the spatial variation in layer thickness, from X-ray computed tomography for use in fuel performance modelling. Simulated irradiation of a demonstration particle found SiC hoop stress values peaking at 315 MPa in tension, significantly in excess of those from previous modelling studies with similar particle aspect ratios. Simulations with representative 2D axisymmetric geometries based on the demonstration particle predicted significantly lower stresses for the same simulated irradiation. 2D radial segments extracted with an arbitrarily oriented polar axis under-predicted the maximum SiC hoop stress by 315-400 MPa, while those extracted with the polar axis passing through the point of maximum SiC hoop stress in the 3D model under-predicted the maximum SiC hoop stress by 165-275 MPa. The 2D model produced using existing methods for generating a 2D flat-spot particle under-predicted the maximum SiC hoop stress by 215 MPa. These findings suggest that existing models may underestimate the stress caused by the asphericity of certain TRISO particle morphologies, and that the current AGR specification may not capture all of the geometric factors that contribute to particle failure probability.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"608 ","pages":"Article 155714"},"PeriodicalIF":2.8,"publicationDate":"2025-02-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143511580","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

The effect of precipitates and alloying elements on γ-Fe (111) surface dissolution corrosion in liquid lead-bismuth eutectic by first-principles study

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

Journal of Nuclear Materials

Pub Date : 2025-02-26 DOI: 10.1016/j.jnucmat.2025.155716

Yufei Li , Runyu Zhou , Tao Gao , Changan Chen

The M₂₃C₆ precipitates play an important role in the corrosion behavior of austenitic stainless steel. Here, by establishing the connection between the surface and the precipitated phase, this work assesses the surface dissolution corrosion by applying the electrode potential. Firstly, the influence of different carbides precipitates on surface dissolution corrosion is compared, which shows that Cr₂₂FeC₆ has the highest electrode potential (+1.68 V), accelerating surface dissolution corrosion. Then, the effect of solute atoms (Pb/Bi/O) on surface dissolution corrosion is studied. It is found that Pb/Bi will promote the dissolution of surface Fe atoms. However, O will strengthen the corrosion resistance of the surface. Simultaneously, the O inhibits the hybridization of the 3d orbital of Fe and the 6p orbital of Bi/Pb, mitigating the corrosion of Pb/Bi on the surface. Lastly, the effects of common alloying elements (Al, Si, and Ni) in austenitic steel on the corrosion of the surface are also investigated to improve surface corrosion resistance. This research attempts to provide a more comprehensive knowledge of the corrosion of iron substrates in ADSs, improving the safety of nuclear energy systems.

{"title":"The effect of precipitates and alloying elements on γ-Fe (111) surface dissolution corrosion in liquid lead-bismuth eutectic by first-principles study","authors":"Yufei Li , Runyu Zhou , Tao Gao , Changan Chen","doi":"10.1016/j.jnucmat.2025.155716","DOIUrl":"10.1016/j.jnucmat.2025.155716","url":null,"abstract":"<div><div>The M<sub>23</sub>C<sub>6</sub> precipitates play an important role in the corrosion behavior of austenitic stainless steel. Here, by establishing the connection between the surface and the precipitated phase, this work assesses the surface dissolution corrosion by applying the electrode potential. Firstly, the influence of different carbides precipitates on surface dissolution corrosion is compared, which shows that Cr<sub>22</sub>FeC<sub>6</sub> has the highest electrode potential (+1.68 V), accelerating surface dissolution corrosion. Then, the effect of solute atoms (Pb/Bi/O) on surface dissolution corrosion is studied. It is found that Pb/Bi will promote the dissolution of surface Fe atoms. However, O will strengthen the corrosion resistance of the surface. Simultaneously, the O inhibits the hybridization of the 3d orbital of Fe and the 6p orbital of Bi/Pb, mitigating the corrosion of Pb/Bi on the surface. Lastly, the effects of common alloying elements (Al, Si, and Ni) in austenitic steel on the corrosion of the surface are also investigated to improve surface corrosion resistance. This research attempts to provide a more comprehensive knowledge of the corrosion of iron substrates in ADSs, improving the safety of nuclear energy systems.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"608 ","pages":"Article 155716"},"PeriodicalIF":2.8,"publicationDate":"2025-02-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143528685","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

Statistical fracture behavior of doped UO2 using a ball-on-ring equibiaxial flexure test method

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

Journal of Nuclear Materials

Pub Date : 2025-02-25 DOI: 10.1016/j.jnucmat.2025.155713

Adrianna E. Lupercio , Tashiema L. Ulrich , Andrew T. Nelson , Brian J. Jaques

Metal oxide dopants, such as titanium and chromium oxides, have garnered considerable attention for their potential to increase grain size (≥ 30 µm) in UO₂ fuel, purportedly enhancing fission gas retention during reactor operation. Fuel performance is significantly impacted by fuel fracture behavior, so it is important to understand the effects of enhanced grain size and dopant content on UO₂ fuel fracture. UO₂ pellets were doped with 0.1 wt% TiO₂ and 0.3 wt% Cr₂O₃ to alter density and grain size. Inductively coupled plasma mass spectroscopy measured dopant levels pre- and post-sintering. X-ray diffraction revealed lattice changes and microstrain via Rietveld refinement. Field emission scanning electron microscopy determined grain sizes of approximately 30 µm for TiO₂ doping and 7 µm for Cr₂O₃ doping. Transverse rupture strength tests were performed on over 30 samples per dataset to obtain characteristic strength and Weibull modulus. Results indicate no statistical difference in fracture strength between 0.1 wt% TiO₂ doped UO₂ and undoped UO₂, while 0.3 wt% Cr₂O₃ doped UO₂ exhibited a 20% decrease in fracture strength. Doped UO₂ samples also showed reduced Weibull modulus compared to undoped UO₂, suggesting increased scatter in fracture strength. This study's findings suggest that titanium and chromium oxide doping in UO₂, regardless of grain size, induce residual stresses, decreasing fracture strength and increasing variability in fracture behavior.

{"title":"Statistical fracture behavior of doped UO2 using a ball-on-ring equibiaxial flexure test method","authors":"Adrianna E. Lupercio , Tashiema L. Ulrich , Andrew T. Nelson , Brian J. Jaques","doi":"10.1016/j.jnucmat.2025.155713","DOIUrl":"10.1016/j.jnucmat.2025.155713","url":null,"abstract":"<div><div>Metal oxide dopants, such as titanium and chromium oxides, have garnered considerable attention for their potential to increase grain size (≥ 30 µm) in UO<sub>2</sub> fuel, purportedly enhancing fission gas retention during reactor operation. Fuel performance is significantly impacted by fuel fracture behavior, so it is important to understand the effects of enhanced grain size and dopant content on UO<sub>2</sub> fuel fracture. UO<sub>2</sub> pellets were doped with 0.1 wt% TiO<sub>2</sub> and 0.3 wt% Cr<sub>2</sub>O<sub>3</sub> to alter density and grain size. Inductively coupled plasma mass spectroscopy measured dopant levels pre- and post-sintering. X-ray diffraction revealed lattice changes and microstrain via Rietveld refinement. Field emission scanning electron microscopy determined grain sizes of approximately 30 µm for TiO<sub>2</sub> doping and 7 µm for Cr<sub>2</sub>O<sub>3</sub> doping. Transverse rupture strength tests were performed on over 30 samples per dataset to obtain characteristic strength and Weibull modulus. Results indicate no statistical difference in fracture strength between 0.1 wt% TiO<sub>2</sub> doped UO<sub>2</sub> and undoped UO<sub>2</sub>, while 0.3 wt% Cr<sub>2</sub>O<sub>3</sub> doped UO<sub>2</sub> exhibited a 20% decrease in fracture strength. Doped UO<sub>2</sub> samples also showed reduced Weibull modulus compared to undoped UO<sub>2</sub>, suggesting increased scatter in fracture strength. This study's findings suggest that titanium and chromium oxide doping in UO<sub>2</sub>, regardless of grain size, induce residual stresses, decreasing fracture strength and increasing variability in fracture behavior.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"608 ","pages":"Article 155713"},"PeriodicalIF":2.8,"publicationDate":"2025-02-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143548941","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

In situ self-ion (Fe+) irradiation of ODS-FeCrAl alloy fuel cladding materials with different Cr contents: The early stages of Cr-rich α’ phase precipitation

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

Journal of Nuclear Materials

Pub Date : 2025-02-22 DOI: 10.1016/j.jnucmat.2025.155706

Hoang Le , Yann de Carlan , David T. Hoelzer , Kan Sakamoto , Per O.Å. Persson , Jonathan A. Hinks , Konstantina Lambrinou

Oxide-dispersion-strengthened FeCrAl (ODS-FeCrAl) alloys are candidate accident-tolerant fuel cladding materials for light water reactors because they demonstrate satisfactory resistance to materials degradation effects such as high-temperature oxidation, radiation-induced swelling, and creep. Their perspective deployment to market is challenged, however, by their inherent susceptibility to irradiation embrittlement caused by the precipitation of the brittle Cr-rich α’ phase at relatively low temperatures (≤475 °C). This work used in situ self-ion irradiation (150 keV Fe⁺) in a transmission electron microscope to elucidate the early stages of Cr-rich α’ phase precipitation in three candidate ODS-FeCrAl alloy fuel cladding materials with different Cr contents (10, 12, and 20 wt.%) and microstructures. The early stages of the process resulting in the precipitation of the Cr-rich α’ phase in these three ODS-FeCrAl alloys under Fe⁺ irradiation were investigated at room temperature and 300 °C up to total fluences of 1.7 × 10¹⁵ ions·cm^-2 (2 dpa) and 3.4 × 10¹⁵ ions·cm^-2 (4 dpa), using three damage dose rates (5 × 10^–5, 3.3 × 10^–4, and 2 × 10^–3 dpa·s^-1). Post-irradiation examination via scanning transmission electron microscopy, energy-dispersive X-ray spectroscopy and electron energy loss spectroscopy suggested that the precipitation of the Cr-rich α’ phase might be promoted by the phase separation of the alloy matrix into Cr-rich and Fe-rich regions. Interestingly, oxygen impurities segregated preferentially in the Cr-rich regions, possibly promoting the radiation-assisted formation of the Cr-rich α’ phase. α’ phase precipitation was more pronounced at room temperature when compared to 300 °C, and it was clearly promoted by the progressive increase in the Cr content of the ODS-FeCrAl alloy.

{"title":"In situ self-ion (Fe+) irradiation of ODS-FeCrAl alloy fuel cladding materials with different Cr contents: The early stages of Cr-rich α’ phase precipitation","authors":"Hoang Le , Yann de Carlan , David T. Hoelzer , Kan Sakamoto , Per O.Å. Persson , Jonathan A. Hinks , Konstantina Lambrinou","doi":"10.1016/j.jnucmat.2025.155706","DOIUrl":"10.1016/j.jnucmat.2025.155706","url":null,"abstract":"<div><div>Oxide-dispersion-strengthened FeCrAl (ODS-FeCrAl) alloys are candidate accident-tolerant fuel cladding materials for light water reactors because they demonstrate satisfactory resistance to materials degradation effects such as high-temperature oxidation, radiation-induced swelling, and creep. Their perspective deployment to market is challenged, however, by their inherent susceptibility to irradiation embrittlement caused by the precipitation of the brittle Cr-rich α’ phase at relatively low temperatures (≤475 °C). This work used <em>in situ</em> self-ion irradiation (150 keV Fe<sup>+</sup>) in a transmission electron microscope to elucidate the early stages of Cr-rich α’ phase precipitation in three candidate ODS-FeCrAl alloy fuel cladding materials with different Cr contents (10, 12, and 20 wt.%) and microstructures. The early stages of the process resulting in the precipitation of the Cr-rich α’ phase in these three ODS-FeCrAl alloys under Fe<sup>+</sup> irradiation were investigated at room temperature and 300 °C up to total fluences of 1.7 × 10<sup>15</sup> ions·cm<sup>-2</sup> (2 dpa) and 3.4 × 10<sup>15</sup> ions·cm<sup>-2</sup> (4 dpa), using three damage dose rates (5 × 10<sup>–5</sup>, 3.3 × 10<sup>–4</sup>, and 2 × 10<sup>–3</sup> dpa·s<sup>-1</sup>). Post-irradiation examination via scanning transmission electron microscopy, energy-dispersive X-ray spectroscopy and electron energy loss spectroscopy suggested that the precipitation of the Cr-rich α’ phase might be promoted by the phase separation of the alloy matrix into Cr-rich and Fe-rich regions. Interestingly, oxygen impurities segregated preferentially in the Cr-rich regions, possibly promoting the radiation-assisted formation of the Cr-rich α’ phase. α’ phase precipitation was more pronounced at room temperature when compared to 300 °C, and it was clearly promoted by the progressive increase in the Cr content of the ODS-FeCrAl alloy.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"608 ","pages":"Article 155706"},"PeriodicalIF":2.8,"publicationDate":"2025-02-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143535245","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

Experimental study on the oxidation behavior of nuclear graphite IG-110

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

Journal of Nuclear Materials

Pub Date : 2025-02-22 DOI: 10.1016/j.jnucmat.2025.155712

Bin Lin , Liang Chen , Guisen Liu , Yao Shen

This study presents an experimental investigation on the oxidation behavior of nuclear graphite IG-110. A gas chromatography-based oxidation test platform was developed to monitor the oxidation processes of IG-110 graphite in different atmospheres. The oxidation tests focused on the presence of H₂O and O₂, with temperatures up to 750 °C and various partial pressures. The graphite consumption was analyzed, and a Boltzmann function was proposed to describe the relationship between graphite consumption and oxygen partial pressure. A graphite consumption prediction model was developed based on this function, which can be utilized to calculate the graphite consumption of CO/CO₂ reaction products in High-Temperature Gas-cooled Reactors (HTGRs).

引用次数: 0

LBE erosion-corrosion behaviors of gelcasted high-speed rotating Ti3AlC2 impeller

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

Journal of Nuclear Materials

Pub Date : 2025-02-21 DOI: 10.1016/j.jnucmat.2025.155708

Zhanchong Zhao , Shibo Shi , Xinxin Gao , Qingsheng Wang , Youyuan Zhang , Yuying Wen , Yang Ling , Xian Zeng , Mei Ma

The MAX phase stands out as one of the highly prospective candidate materials for the impeller of the lead-bismuth fast reactor nuclear main pump. In order to determine the compatibility between the high-speed rotating Ti₃AlC₂ ceramic impeller and the liquid lead-bismuth eutectic (LBE) alloy, the corrosion behavior was comprehensively investigated within the LBE environment. In this study, the erosion-corrosion behavior of a high-speed rotating Ti₃AlC₂ ceramic impeller in LBE at 550 °C for 600 h was systematically investigated. The research results indicate that the Ti₃AlC₂ impeller maintained its structural integrity without macroscopic fractures after dynamic corrosion. Weight evaluation shows that during the corrosion test, the degradation of its physical or chemical properties was negligible. Surface analysis revealed the formation of a corrosion layer primarily composed of TiC and amorphous carbon, with PbO adhering to the surface as a protective barrier, effectively mitigating Ti and Al losses. Comparative analysis confirmed the superior adhesiveness of PbO over TiO₂ and Al₂O₃. The impeller maintained its mechanical performance, experiencing only a minor weight gain of 0.02 wt.% during the test. Vibration analysis confirmed operational stability, with a maximum stress of 10.76 MPa and a rotational frequency of 16.7 Hz, well below the first-order resonance frequency of 4997.7 Hz. This study furnishes crucial insights into the corrosion characteristics of the MAX phase and presents significant data. It is anticipated to provide a valuable reference for the initial application of MAX-phase ceramic impellers in advanced nuclear systems.

{"title":"LBE erosion-corrosion behaviors of gelcasted high-speed rotating Ti3AlC2 impeller","authors":"Zhanchong Zhao , Shibo Shi , Xinxin Gao , Qingsheng Wang , Youyuan Zhang , Yuying Wen , Yang Ling , Xian Zeng , Mei Ma","doi":"10.1016/j.jnucmat.2025.155708","DOIUrl":"10.1016/j.jnucmat.2025.155708","url":null,"abstract":"<div><div>The MAX phase stands out as one of the highly prospective candidate materials for the impeller of the lead-bismuth fast reactor nuclear main pump. In order to determine the compatibility between the high-speed rotating Ti<sub>3</sub>AlC<sub>2</sub> ceramic impeller and the liquid lead-bismuth eutectic (LBE) alloy, the corrosion behavior was comprehensively investigated within the LBE environment. In this study, the erosion-corrosion behavior of a high-speed rotating Ti<sub>3</sub>AlC<sub>2</sub> ceramic impeller in LBE at 550 °C for 600 h was systematically investigated. The research results indicate that the Ti<sub>3</sub>AlC<sub>2</sub> impeller maintained its structural integrity without macroscopic fractures after dynamic corrosion. Weight evaluation shows that during the corrosion test, the degradation of its physical or chemical properties was negligible. Surface analysis revealed the formation of a corrosion layer primarily composed of TiC and amorphous carbon, with PbO adhering to the surface as a protective barrier, effectively mitigating Ti and Al losses. Comparative analysis confirmed the superior adhesiveness of PbO over TiO<sub>2</sub> and Al<sub>2</sub>O<sub>3</sub>. The impeller maintained its mechanical performance, experiencing only a minor weight gain of 0.02 wt.% during the test. Vibration analysis confirmed operational stability, with a maximum stress of 10.76 MPa and a rotational frequency of 16.7 Hz, well below the first-order resonance frequency of 4997.7 Hz. This study furnishes crucial insights into the corrosion characteristics of the MAX phase and presents significant data. It is anticipated to provide a valuable reference for the initial application of MAX-phase ceramic impellers in advanced nuclear systems.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"607 ","pages":"Article 155708"},"PeriodicalIF":2.8,"publicationDate":"2025-02-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143487305","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

X-ray computed tomography of deconsolidated TRISO Particles from the AGR-5/6/7 irradiation experiment capsule 1 Compact

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

Journal of Nuclear Materials

Pub Date : 2025-02-19 DOI: 10.1016/j.jnucmat.2025.155704

Rahul Reddy Kancharla , William C. Chuirazzi , Joshua J. Kane , John D. Stempien , Cameron Howard , Swapnil Morankar , Miles T. Cook , Quintin D. Harris

X-ray Computed Tomography (XCT) was performed on TRISO particles from fuel Compact 1-7-9 from Capsule 1 of the AGR-5/6/7 irradiation experiment. Compact 1–7–9 was in the vicinity of overheated thermocouples that released transition metals and caused significant damage and failure of TRISO particles within regions of Capsule 1. Four particles, Particles A–D, were non-destructively examined in three-dimensions (3D). Particle B was found to have degraded silicon carbide (SiC) and significant amounts of Ni, actinides, and fission products that contributed to this degradation. The nickel (Ni) showed some evidence of penetration into Particle B from the exterior surface of the silicon carbide layer. The kernels of Particles A, C, and D were analyzed for their size, shape, and observed porosity. A density gradient was observed in the kernels, which may be nascent kernel migration observed in fuel with higher irradiation temperatures. In addition to the kernels, Particles A, C, and D possessed large clusters of dense material that were grouped primarily to one side of the particle located at the buffer/inner pyrolytic carbon interface. These clusters were attributed to nickel migration into the particle interiors, which was confirmed with scanning electron microscopy (SEM) and electron dispersive X-ray spectroscopy (EDS).

{"title":"X-ray computed tomography of deconsolidated TRISO Particles from the AGR-5/6/7 irradiation experiment capsule 1 Compact","authors":"Rahul Reddy Kancharla , William C. Chuirazzi , Joshua J. Kane , John D. Stempien , Cameron Howard , Swapnil Morankar , Miles T. Cook , Quintin D. Harris","doi":"10.1016/j.jnucmat.2025.155704","DOIUrl":"10.1016/j.jnucmat.2025.155704","url":null,"abstract":"<div><div>X-ray Computed Tomography (XCT) was performed on TRISO particles from fuel Compact 1-7-9 from Capsule 1 of the AGR-5/6/7 irradiation experiment. Compact 1–7–9 was in the vicinity of overheated thermocouples that released transition metals and caused significant damage and failure of TRISO particles within regions of Capsule 1. Four particles, Particles A–D, were non-destructively examined in three-dimensions (3D). Particle B was found to have degraded silicon carbide (SiC) and significant amounts of Ni, actinides, and fission products that contributed to this degradation. The nickel (Ni) showed some evidence of penetration into Particle B from the exterior surface of the silicon carbide layer. The kernels of Particles A, C, and D were analyzed for their size, shape, and observed porosity. A density gradient was observed in the kernels, which may be nascent kernel migration observed in fuel with higher irradiation temperatures. In addition to the kernels, Particles A, C, and D possessed large clusters of dense material that were grouped primarily to one side of the particle located at the buffer/inner pyrolytic carbon interface. These clusters were attributed to nickel migration into the particle interiors, which was confirmed with scanning electron microscopy (SEM) and electron dispersive X-ray spectroscopy (EDS).</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"607 ","pages":"Article 155704"},"PeriodicalIF":2.8,"publicationDate":"2025-02-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143478884","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

Reaction behavior of molten 316L stainless steel with B4C at 1450℃ during a core melt accident of BWR

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

Journal of Nuclear Materials

Pub Date : 2025-02-18 DOI: 10.1016/j.jnucmat.2025.155705

Tatsuya Kanno , Takayuki Iwama , Takumi Sato , Ayumi Itoh , Yuji Nagae , Ryo Inoue , Shigeru Ueda

In the severe accident at the Fukushima Daiichi Nuclear Power Station, molten core material migrated downward and generated fuel debris at the bottom of the containment vessel and other locations. The composition of the generated melt changes as it moves because it reacts with structural materials. The composition of fuel debris is determined by the initial melt composition and reactions during migration. In this study, the reaction rate between molten 316 L stainless steel and B₄C at 1450 °C was experimentally investigated, assuming that control rods would melt and stainless steel-based melts would stagnate in the lower plenum in boiling water reactors. The effect of H₂O gas on the reaction between the molten metal and B₄C was investigated, and an estimation of the rate-limiting process of the reaction and the rate of B₄C was carried out. The rate-limiting process of the reaction changed from the dissolution reaction of B₄C to the diffusion of alloying elements and the precipitation of carbon as the reaction proceeded. Including these rate-limiting processes, an overall dissolution rate of B₄C melting in the B₄C/molten stainless steel reaction at 1450 °C was obtained.

{"title":"Reaction behavior of molten 316L stainless steel with B4C at 1450℃ during a core melt accident of BWR","authors":"Tatsuya Kanno , Takayuki Iwama , Takumi Sato , Ayumi Itoh , Yuji Nagae , Ryo Inoue , Shigeru Ueda","doi":"10.1016/j.jnucmat.2025.155705","DOIUrl":"10.1016/j.jnucmat.2025.155705","url":null,"abstract":"<div><div>In the severe accident at the Fukushima Daiichi Nuclear Power Station, molten core material migrated downward and generated fuel debris at the bottom of the containment vessel and other locations. The composition of the generated melt changes as it moves because it reacts with structural materials. The composition of fuel debris is determined by the initial melt composition and reactions during migration. In this study, the reaction rate between molten 316 L stainless steel and B<sub>4</sub>C at 1450 °C was experimentally investigated, assuming that control rods would melt and stainless steel-based melts would stagnate in the lower plenum in boiling water reactors. The effect of H<sub>2</sub>O gas on the reaction between the molten metal and B<sub>4</sub>C was investigated, and an estimation of the rate-limiting process of the reaction and the rate of B₄C was carried out. The rate-limiting process of the reaction changed from the dissolution reaction of B<sub>4</sub>C to the diffusion of alloying elements and the precipitation of carbon as the reaction proceeded. Including these rate-limiting processes, an overall dissolution rate of B₄C melting in the B₄C/molten stainless steel reaction at 1450 °C was obtained.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"607 ","pages":"Article 155705"},"PeriodicalIF":2.8,"publicationDate":"2025-02-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143487304","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

Development of SiC encapsulation for thulium oxide targets as potential nuclear batteries

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

Journal of Nuclear Materials

Pub Date : 2025-02-17 DOI: 10.1016/j.jnucmat.2025.155700

Brandon Shaver , Kip Wheeler , Caen Ang

The design, fabrication and properties of encapsulations for radioisotope fuel targets (¹⁶⁹Tm) were assessed. Elemental analysis and phase identification by X-ray Fluorescence and X-ray Diffraction showed a SiC encapsulation of a thulium oxide ceramic. Microhardness testing of the encapsulation showed a relatively consistent hardness of 32.2 ± 4.7 GPa across the vertices and walls, with a mid-plane join hardness of 30.1 ± 5.4 GPa. Manufacturing tolerances followed a normal distribution with a standard deviation (σ) of σ _OD = ± 0.014 mm and σ_ID = ± 0.037 mm for Tm₂O₃ and SiC. The design of the target emphasizes robust, radiation-tolerant, high-strength SiC, but heat transfer is likely axially asymmetric because contact between Tm₂O₃-SiC is limited. Tm₂O₃-SiC chemical compatibility testing was investigated, indicating a possible reaction-limited process below 1873 K, and a possible diffusion-limited process above 1873 K. At higher temperatures, thulium containment for this concept is limited by chemical compatibility.

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引用次数: 0

Mechanistic insights into spinodal L12 nanostructures in mitigating radiation defect growth in Al0.5Cr0.9FeNi2.5V0.2 high-entropy alloys

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

Journal of Nuclear Materials

Pub Date : 2025-02-16 DOI: 10.1016/j.jnucmat.2025.155703

Zhengxiong Su, Jianqiang Wang, Jinxue Yang, Ping Zhang, Rui Gao, Chenyang Lu

High-entropy alloys with high-content L1₂ nanoprecipitates formed through phase separation have recently demonstrated outstanding mechanical properties across a wide temperature range, making them suitable structural materials for advanced nuclear systems. This study investigates the irradiation response of a high-entropy alloy composed of Al_0.5Cr_0.9FeNi_2.5V_0.2 with varying volumes of L1₂ nanostructures. High-temperature He ion irradiation was performed, and its effects on defect evolution were analyzed using transmission electron microscopy and nanoindentation. The results show that the spinodal order-disorder L1₂ network structure effectively suppresses irradiation-induced defects, as evidenced by reduced dislocation loop size, lower He bubble swelling, and decreased irradiation hardening in alloys with higher L1₂ volume fractions. This is primarily because the L1₂ structure, with its low-misfit coherent interface, undergoes the reversible order-disorder transition that reduces early irradiation point defects and suppresses defect nucleation and growth. Furthermore, the spinodal order-disorder L1₂ nanostructure impedes defect cluster movement by providing diffuse obstacles and forming antiphase boundaries, thereby slowing the growth of defects and He bubbles. This work provides an alloy design strategy to improve irradiation tolerance by exploiting the self-healing and structural complexity of the L1₂ structure.

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