As promising nuclear structural materials, high-entropy alloys have attracted extensive attentions due to their excellent anti-irradiation properties. In this work, the crystal-orientation-dependence of irradiation damage in equiatomic CoCrFeNiMn alloy, subjected to 2.5 MeV Fe²⁺ irradiation up to 15 dpa at 500 °C, was investigated using transmission electron microscopy (TEM). The 〈001〉-oriented grain exhibited the worst irradiation resistance, as evidenced by the densest 1/2〈110〉 perfect dislocations, the largest faulted loops and the greatest void swelling among all the damaged orientations. The 〈011〉 grain showed the lowest void swelling. Furthermore, the 〈111〉 grain exhibited faulted loops that were smaller in size and exhibited a higher density. Possible contributions to the orientation-dependent radiation damage of CoCrFeNiMn were discussed. The minimal damage observed in 〈011〉 grain can be attributed to its highest degree of channeling. It is also crucial to consider other contributing factors as well for 〈001〉 and 〈111〉 grains. This study underscores the importance of accounting for orientation when assessing the irradiation damage behavior of polycrystalline SP-CSAs with substantial grain sizes.
{"title":"Crystal-orientation-dependence of irradiation damage in CoCrFeNiMn alloy under heavy ion irradiation at 500°C","authors":"Fei Zhu, Feng Zhou, Qiang Zhang, Jinhong Chen, Jiaxin Wu, Ligang Song, Xianfeng Ma","doi":"10.1016/j.jnucmat.2025.155729","DOIUrl":"10.1016/j.jnucmat.2025.155729","url":null,"abstract":"<div><div>As promising nuclear structural materials, high-entropy alloys have attracted extensive attentions due to their excellent anti-irradiation properties. In this work, the crystal-orientation-dependence of irradiation damage in equiatomic CoCrFeNiMn alloy, subjected to 2.5 MeV Fe²⁺ irradiation up to 15 dpa at 500 °C, was investigated using transmission electron microscopy (TEM). The 〈001〉-oriented grain exhibited the worst irradiation resistance, as evidenced by the densest 1/2〈110〉 perfect dislocations, the largest faulted loops and the greatest void swelling among all the damaged orientations. The 〈011〉 grain showed the lowest void swelling. Furthermore, the 〈111〉 grain exhibited faulted loops that were smaller in size and exhibited a higher density. Possible contributions to the orientation-dependent radiation damage of CoCrFeNiMn were discussed. The minimal damage observed in 〈011〉 grain can be attributed to its highest degree of channeling. It is also crucial to consider other contributing factors as well for 〈001〉 and 〈111〉 grains. This study underscores the importance of accounting for orientation when assessing the irradiation damage behavior of polycrystalline SP-CSAs with substantial grain sizes.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"608 ","pages":"Article 155729"},"PeriodicalIF":2.8,"publicationDate":"2025-03-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143577136","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2025-03-04DOI: 10.1016/j.jnucmat.2025.155728
Suming Chen , Yangchun Chen , Rongjian Pan , Xichuan Liao , Rongyang Qiu , Long Guo , Zhixiao Liu , Huiqiu Deng
Tungsten-based high-entropy alloys have demonstrated promising performance as materials for nuclear fusion. Understanding typical helium behavior in these alloys is crucial for assessing their resistance to helium ion irradiation and underlying mechanisms. In this study, we developed a W-Ta-Cr-V-He five-element interatomic potential and used it to examine helium behavior in the W38Ta36Cr15V11 alloy through molecular dynamics simulations. Our results reveal several notable differences between W38Ta36Cr15V11 and pure tungsten. Specifically, the formation and binding energies of helium clusters and helium-vacancy clusters in W38Ta36Cr15V11 are significantly lower than in pure tungsten, indicating reduced binding ability for helium atoms and a weaker self-trapping effect. Furthermore, the alloy exhibits a significantly higher diffusion energy barrier for a single interstitial helium atom, resulting in decreased helium mobility and further reducing the tendency for helium cluster formation. The study also highlights distinct mechanisms of helium bubble growth: in W38Ta36Cr15V11, helium clusters lead to the expulsion of interstitial atoms without forming dislocation loops, whereas in pure tungsten, dislocation loop emission accompanies helium bubble growth. Temperature-dependent simulations show that helium bubble nucleation is notably suppressed in W38Ta36Cr15V11 compared to pure tungsten, with weaker clustering of helium atoms observed at various temperatures. The developed W-Ta-Cr-V-He potential and the resulting data offer valuable insights into helium behavior in tungsten-based high-entropy alloys.
{"title":"Helium behavior in W-Ta-Cr-V high-entropy alloy: An interatomic potential and molecular dynamics simulations","authors":"Suming Chen , Yangchun Chen , Rongjian Pan , Xichuan Liao , Rongyang Qiu , Long Guo , Zhixiao Liu , Huiqiu Deng","doi":"10.1016/j.jnucmat.2025.155728","DOIUrl":"10.1016/j.jnucmat.2025.155728","url":null,"abstract":"<div><div>Tungsten-based high-entropy alloys have demonstrated promising performance as materials for nuclear fusion. Understanding typical helium behavior in these alloys is crucial for assessing their resistance to helium ion irradiation and underlying mechanisms. In this study, we developed a W-Ta-Cr-V-He five-element interatomic potential and used it to examine helium behavior in the W<sub>38</sub>Ta<sub>36</sub>Cr<sub>15</sub>V<sub>11</sub> alloy through molecular dynamics simulations. Our results reveal several notable differences between W<sub>38</sub>Ta<sub>36</sub>Cr<sub>15</sub>V<sub>11</sub> and pure tungsten. Specifically, the formation and binding energies of helium clusters and helium-vacancy clusters in W<sub>38</sub>Ta<sub>36</sub>Cr<sub>15</sub>V<sub>11</sub> are significantly lower than in pure tungsten, indicating reduced binding ability for helium atoms and a weaker self-trapping effect. Furthermore, the alloy exhibits a significantly higher diffusion energy barrier for a single interstitial helium atom, resulting in decreased helium mobility and further reducing the tendency for helium cluster formation. The study also highlights distinct mechanisms of helium bubble growth: in W<sub>38</sub>Ta<sub>36</sub>Cr<sub>15</sub>V<sub>11</sub>, helium clusters lead to the expulsion of interstitial atoms without forming dislocation loops, whereas in pure tungsten, dislocation loop emission accompanies helium bubble growth. Temperature-dependent simulations show that helium bubble nucleation is notably suppressed in W<sub>38</sub>Ta<sub>36</sub>Cr<sub>15</sub>V<sub>11</sub> compared to pure tungsten, with weaker clustering of helium atoms observed at various temperatures. The developed W-Ta-Cr-V-He potential and the resulting data offer valuable insights into helium behavior in tungsten-based high-entropy alloys.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"608 ","pages":"Article 155728"},"PeriodicalIF":2.8,"publicationDate":"2025-03-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143562626","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2025-03-01DOI: 10.1016/j.jnucmat.2025.155719
K. Abbas , M. Chebbi , B. Azambre , C. Monsanglant-Louvet , B. Marcillaud , A. Roynette
The efficient capture of radioactive methyl iodide (CH3131I) is a critical issue for nuclear safety and radioprotection. Co-impregnated activated carbons (AC), with triethylenediamine (TEDA) and potassium iodide (K127I), are widely employed for this purpose. However, the specific role of KI in CH3131I retention through isotopic exchange reaction remains poorly understood. This study provides groundbreaking insights by systematically investigating the retention behavior of KI/AC versus non-impregnated activated carbons (NI AC) under different operating conditions. Advanced characterization techniques, including N2 porosimetry, high-resolution transmission electron microscopy (HRTEM), and H2O adsorption isotherms, were first employed to elucidate the structural and chemical properties of the adsorbents. Subsequently, CH3131I retention tests were conducted by measuring the Decontamination Factors (DF) at various configurations covering a broad range of relative humidities (RH) (20–90 %), temperatures (20–96 °C), residence times (0.125–0.5 s) and elution times (1–18 h). Results reveal that while NI AC exhibits a drastic performance decline at high RH attributable to water physisorption, KI/AC demonstrates enhanced retention, counterbalancing moisture effects via isotopic exchange. Furthermore, elevated temperatures significantly amplify DF for KI/AC, unveiling for the first time the thermally activated nature of the isotopic exchange mechanism. Prolonged residence time further enhance performance for KI/AC compared to NI AC, suggesting multiple mechanistic steps in isotopic exchange reaction. Consequently, a detailed mechanism for this reaction has been proposed.
This work advances the understanding of CH3131I capture mechanisms ensuring improved performance under diverse nuclear safety scenarios.
{"title":"Unravelling the potential of iodine isotopic exchange in CH3131I capture by K127I-impregnated activated carbons","authors":"K. Abbas , M. Chebbi , B. Azambre , C. Monsanglant-Louvet , B. Marcillaud , A. Roynette","doi":"10.1016/j.jnucmat.2025.155719","DOIUrl":"10.1016/j.jnucmat.2025.155719","url":null,"abstract":"<div><div>The efficient capture of radioactive methyl iodide (CH<sub>3</sub><sup>131</sup>I) is a critical issue for nuclear safety and radioprotection. Co-impregnated activated carbons (AC), with triethylenediamine (TEDA) and potassium iodide (K<sup>127</sup>I), are widely employed for this purpose. However, the specific role of KI in CH<sub>3</sub><sup>131</sup>I retention through isotopic exchange reaction remains poorly understood. This study provides groundbreaking insights by systematically investigating the retention behavior of KI/AC <em>versus</em> non-impregnated activated carbons (NI AC) under different operating conditions. Advanced characterization techniques, including N<sub>2</sub> porosimetry, high-resolution transmission electron microscopy (HRTEM), and H<sub>2</sub>O adsorption isotherms, were first employed to elucidate the structural and chemical properties of the adsorbents. Subsequently, CH<sub>3</sub><sup>131</sup>I retention tests were conducted by measuring the Decontamination Factors (DF) at various configurations covering a broad range of relative humidities (RH) (20–90 %), temperatures (20–96 °C), residence times (0.125–0.5 s) and elution times (1–18 h). Results reveal that while NI AC exhibits a drastic performance decline at high RH attributable to water physisorption, KI/AC demonstrates enhanced retention, counterbalancing moisture effects <em>via</em> isotopic exchange. Furthermore, elevated temperatures significantly amplify DF for KI/AC, unveiling for the first time the thermally activated nature of the isotopic exchange mechanism. Prolonged residence time further enhance performance for KI/AC compared to NI AC, suggesting multiple mechanistic steps in isotopic exchange reaction. Consequently, a detailed mechanism for this reaction has been proposed.</div><div>This work advances the understanding of CH<sub>3</sub><sup>131</sup>I capture mechanisms ensuring improved performance under diverse nuclear safety scenarios.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"608 ","pages":"Article 155719"},"PeriodicalIF":2.8,"publicationDate":"2025-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143548946","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2025-03-01DOI: 10.1016/j.jnucmat.2025.155722
Yijie Tang , Jin Wang , Junxia Wang , Long Kang , Tongming Zhang , Jun Li , Yan Wang , Xusheng Li , Yanping Yang
In this work, MgONd2(Zr1-xCex)2O7 (M-NZCx) composite ceramics, a potential matrix, used for inert matrix fuel were subjected to 3.5 MeV Kr10+ irradiation with fluences of 1 × 1014–5 × 1015 ions/cm2. The phase transition from pyrochlore (P) to defect fluorite (F) structures (P-F transition) in NZCx phase, as well as the lattice expansion, amorphization and strain accumulation in both MgO and pyrochlore phases in post-irradiated M-NZCx composite ceramics were systematically investigated. Specifically, the lattice expansion ratios (Rs) of different phases in M-NZCx samples irradiated with the same fluence of Kr10+ ions were ranged as follows, RsM < RsP < RsF. The further research on irradiation response of different phases in typical M-NZC0.3 sample indicated that MgO exhibited superior irradiation stability to pyrochlore phase. Additionally, the strain accumulation induced by Kr10+ bombardment was observed generally in both MgO and NZC0.3 pyrochlore phases, and the defect-induced strain in MgO crystalline was more pronounced than NZC0.3 pyrochlore phase. Interestingly, the strain accumulation resulting from Kr10+ irradiation was observed to be preferentially oriented along directions with lower atomic density. This study might provide a new perspective for understanding the irradiation stability of MgONd2Zr2O7 based composite ceramics in elastic collision cascade by krypton ions.
{"title":"Kr10+ irradiation stability and strain accumulation of MgONd2(Zr1-xCex)2O7 composite ceramics for inert matrix fuel","authors":"Yijie Tang , Jin Wang , Junxia Wang , Long Kang , Tongming Zhang , Jun Li , Yan Wang , Xusheng Li , Yanping Yang","doi":"10.1016/j.jnucmat.2025.155722","DOIUrl":"10.1016/j.jnucmat.2025.155722","url":null,"abstract":"<div><div>In this work, MgO<img>Nd<sub>2</sub>(Zr<sub>1-</sub><em><sub>x</sub></em>Ce<em><sub>x</sub></em>)<sub>2</sub>O<sub>7</sub> (M-NZC<em><sub>x</sub></em>) composite ceramics, a potential matrix, used for inert matrix fuel were subjected to 3.5 MeV Kr<sup>10+</sup> irradiation with fluences of 1 × 10<sup>14</sup>–5 × 10<sup>15</sup> ions/cm<sup>2</sup>. The phase transition from pyrochlore (P) to defect fluorite (F) structures (P-F transition) in NZC<em><sub>x</sub></em> phase, as well as the lattice expansion, amorphization and strain accumulation in both MgO and pyrochlore phases in post-irradiated M-NZC<em><sub>x</sub></em> composite ceramics were systematically investigated. Specifically, the lattice expansion ratios (<em>Rs</em>) of different phases in M-NZC<em><sub>x</sub></em> samples irradiated with the same fluence of Kr<sup>10+</sup> ions were ranged as follows, <em>Rs</em><sub>M</sub> < <em>Rs</em><sub>P</sub> < <em>Rs</em><sub>F</sub>. The further research on irradiation response of different phases in typical M-NZC<sub>0.3</sub> sample indicated that MgO exhibited superior irradiation stability to pyrochlore phase. Additionally, the strain accumulation induced by Kr<sup>10+</sup> bombardment was observed generally in both MgO and NZC<sub>0.3</sub> pyrochlore phases, and the defect-induced strain in MgO crystalline was more pronounced than NZC<sub>0.3</sub> pyrochlore phase. Interestingly, the strain accumulation resulting from Kr<sup>10+</sup> irradiation was observed to be preferentially oriented along directions with lower atomic density. This study might provide a new perspective for understanding the irradiation stability of MgO<img>Nd<sub>2</sub>Zr<sub>2</sub>O<sub>7</sub> based composite ceramics in elastic collision cascade by krypton ions.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"608 ","pages":"Article 155722"},"PeriodicalIF":2.8,"publicationDate":"2025-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143548944","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}
Austenitic stainless steels utilized in-core components of pressurized water reactors are prone to radiation-induced segregation, which leads to the degradation of microstructure and mechanical properties. To improve irradiation resistance, one possible solution is to increase the number density of point defect sinks, such as grain boundaries. For this purpose, ultrafine-grained or nanostructured microstructures are recommended due to their high density of grain boundaries. This paper investigates the microstructural changes in ultrafine-grained 316 austenitic stainless steel exposed to neutron radiation up to 3.9 dpa in irradiation conditions representative of light water reactors. The microstructure at different length scales was analyzed using electron backscattered diffraction, transmission electron microscopy, and atom probe tomography before and after neutron irradiation. The study compares its findings with those of existing literature on coarse-grained austenitic stainless steels to evaluate the benefit of ultrafine-grained 316 austenitic stainless steels regarding irradiation ageing in representative conditions of light water reactors.
{"title":"Microstructural evolution of neutron irradiated ultrafine-grained austenitic stainless steel","authors":"Frederic Habiyaremye , Solène Rouland , Bertrand Radiguet , Fabien Cuvilly , Benjamin Klaes , Benoit Tanguy , Joël Malaplate , Christophe Domain , Diogo Goncalves , Marina M. Abramova , Nariman A. Enikeev , Xavier Sauvage , Auriane Etienne","doi":"10.1016/j.jnucmat.2025.155710","DOIUrl":"10.1016/j.jnucmat.2025.155710","url":null,"abstract":"<div><div>Austenitic stainless steels utilized in-core components of pressurized water reactors are prone to radiation-induced segregation, which leads to the degradation of microstructure and mechanical properties. To improve irradiation resistance, one possible solution is to increase the number density of point defect sinks, such as grain boundaries. For this purpose, ultrafine-grained or nanostructured microstructures are recommended due to their high density of grain boundaries. This paper investigates the microstructural changes in ultrafine-grained 316 austenitic stainless steel exposed to neutron radiation up to 3.9 dpa in irradiation conditions representative of light water reactors. The microstructure at different length scales was analyzed using electron backscattered diffraction, transmission electron microscopy, and atom probe tomography before and after neutron irradiation. The study compares its findings with those of existing literature on coarse-grained austenitic stainless steels to evaluate the benefit of ultrafine-grained 316 austenitic stainless steels regarding irradiation ageing in representative conditions of light water reactors.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"607 ","pages":"Article 155710"},"PeriodicalIF":2.8,"publicationDate":"2025-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143512334","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}
Pub Date : 2025-03-01DOI: 10.1016/j.jnucmat.2025.155711
Luning Chen, Jing Zhang, Shurong Ding
SiC/SiC composites are one of the promising engineering materials for nuclear applications. Anisotropic swelling deformations were observed in these materials during irradiation, and the underlying mechanisms should be deeply understood. In this study, a numerical simulation method is developed to predict the irradiation-induced deformations of the as-fabricated SiC/SiC composites. An emphasis is given to the generation of an RVE (Representative Volume Element) model with a pre-existing pore and the assumed residual stress field. Besides, the thermo-mechanical constitutive relations and stress update algorithms for the solid skeleton of porous SiC/SiC composites are developed with their irradiation effects considered comprehensively. Based on the homogenization theory, the calculation models to obtain the macroscopic swelling strains of porous SiC/SiC composites are developed. The good agreements between the predictions and the post-irradiation data of anisotropic swelling validate the effectiveness of the developed models and simulation methods. Research findings indicate that the irradiation creep deformations due to the existing residual stresses and high transient creep rate coefficients lead to the through-thickness size shrinkage of the pre-existing pores, which possibly becomes the dominant mechanism of the negative linear swelling of the SiC/SiC sample during the initial irradiation stage. The effects of the initial residual stress fields and the elastic constitutive relations on the anisotropic irradiation swelling behaviors are investigated. This study lays a foundation for the advanced manufacture of the SiC/SiC composites and the based multi-layer cladding tubes.
{"title":"Modeling and analysis for the anisotropic irradiation swelling of porous SiC/SiC composites","authors":"Luning Chen, Jing Zhang, Shurong Ding","doi":"10.1016/j.jnucmat.2025.155711","DOIUrl":"10.1016/j.jnucmat.2025.155711","url":null,"abstract":"<div><div>SiC/SiC composites are one of the promising engineering materials for nuclear applications. Anisotropic swelling deformations were observed in these materials during irradiation, and the underlying mechanisms should be deeply understood. In this study, a numerical simulation method is developed to predict the irradiation-induced deformations of the as-fabricated SiC/SiC composites. An emphasis is given to the generation of an RVE (Representative Volume Element) model with a pre-existing pore and the assumed residual stress field. Besides, the thermo-mechanical constitutive relations and stress update algorithms for the solid skeleton of porous SiC/SiC composites are developed with their irradiation effects considered comprehensively. Based on the homogenization theory, the calculation models to obtain the macroscopic swelling strains of porous SiC/SiC composites are developed. The good agreements between the predictions and the post-irradiation data of anisotropic swelling validate the effectiveness of the developed models and simulation methods. Research findings indicate that the irradiation creep deformations due to the existing residual stresses and high transient creep rate coefficients lead to the through-thickness size shrinkage of the pre-existing pores, which possibly becomes the dominant mechanism of the negative linear swelling of the SiC/SiC sample during the initial irradiation stage. The effects of the initial residual stress fields and the elastic constitutive relations on the anisotropic irradiation swelling behaviors are investigated. This study lays a foundation for the advanced manufacture of the SiC/SiC composites and the based multi-layer cladding tubes.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"607 ","pages":"Article 155711"},"PeriodicalIF":2.8,"publicationDate":"2025-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143512330","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2025-03-01DOI: 10.1016/j.jnucmat.2025.155699
Wulin Song , Xue Han , Huanlin Cheng, Qi Tang, Huacai Wang, Songtao Ji
A comprehensive characterization of the Fuel Cladding Chemical Interaction (FCCI) layer in a Zr-1Nb alloy with a burnup of 41GWd·tU−1 has been performed utilizing primary techniques including Optical Microscopy (OM), Transmission Electron Microscopy (TEM), and Transmission Kikuchi Diffraction (TKD). The results indicate that the FCCI layer characterized in this study is mainly composed of tetragonal zirconia on both the cladding and fuel sides, with monoclinic zirconia in between. Additionally, the strong fiber texture in monoclinic and tetragonal zirconia aligns well with that in the water-side oxide film. The orientation relationships between α-Zr, monoclinic zirconia and tetragonal zirconia are (011) α-Zr || (010) m-ZrO2 and (10)m-ZrO2 || (100) t-ZrO2. It appears that the dominant force for texture development in the FCCI formed on this alloy is the α-Zr to m-ZrO2 and m-ZrO2 to t-ZrO2 transformation stress which is independent with metal substrate orientation.
{"title":"Structure of the fuel-cladding chemical interaction (FCCI) layer of a high burnup Zr-1Nb nuclear fuel cladding","authors":"Wulin Song , Xue Han , Huanlin Cheng, Qi Tang, Huacai Wang, Songtao Ji","doi":"10.1016/j.jnucmat.2025.155699","DOIUrl":"10.1016/j.jnucmat.2025.155699","url":null,"abstract":"<div><div>A comprehensive characterization of the Fuel Cladding Chemical Interaction (FCCI) layer in a Zr-1Nb alloy with a burnup of 41GWd·tU<sup>−1</sup> has been performed utilizing primary techniques including Optical Microscopy (OM), Transmission Electron Microscopy (TEM), and Transmission Kikuchi Diffraction (TKD). The results indicate that the FCCI layer characterized in this study is mainly composed of tetragonal zirconia on both the cladding and fuel sides, with monoclinic zirconia in between. Additionally, the strong fiber texture in monoclinic and tetragonal zirconia aligns well with that in the water-side oxide film. The orientation relationships between α-Zr, monoclinic zirconia and tetragonal zirconia are (<span><math><mover><mn>1</mn><mo>¯</mo></mover></math></span>011) <sub>α-Zr</sub> || (010) <sub>m-ZrO2</sub> and (10<span><math><mover><mn>1</mn><mo>¯</mo></mover></math></span>)<sub>m-ZrO2</sub> || (100) <sub>t-ZrO2</sub>. It appears that the dominant force for texture development in the FCCI formed on this alloy is the α-Zr to m-ZrO<sub>2</sub> and m-ZrO<sub>2</sub> to t-ZrO<sub>2</sub> transformation stress which is independent with metal substrate orientation.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"607 ","pages":"Article 155699"},"PeriodicalIF":2.8,"publicationDate":"2025-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143512335","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2025-03-01DOI: 10.1016/j.jnucmat.2025.155723
Haokun Wang , Shichao Liu , Yuanming Li , Wei Li , Junmei Wu
Silicon Carbide (SiC) is considered a promising candidate for Accident-Tolerant Fuel cladding materials in nuclear reactors. However, existing literature often oversimplifies the heterogeneous geometric characteristics of the braided layers in SiC cladding. This paper presents a detailed modeling of the thermo-mechanical behavior of SiC cladding under light water reactor (LWR) conditions, with a focus on the braiding structure. The yarn, composed of SiC fibers and matrix, is treated as a homogenized orthogonal anisotropic material, and the braiding structure is constructed based on the parametric equations describing yarn paths. The interlayer damage between the yarns and the matrix is modeled using the cohesive zone method. The thermal-mechanical performance of the SiCf/SiC cladding during reactor startup, power operation and reactor shutdown is evaluated. The results confirm a significant increase in the tensile stress of the braided layer during reactor shutdown. Varying the anisotropic swelling of yarns only have slight effect on the cladding stress. Furthermore, the impacts of braiding patterns, braiding angles and fuel rod gap pressure are also investigated. The findings contribute to a more realistic assessment of SiC composite cladding performance, potentially informing future cladding design and fuel safety assessment.
{"title":"Modeling of the thermomechanical behavior of braided SiCf/SiC composite cladding tube during irradiation","authors":"Haokun Wang , Shichao Liu , Yuanming Li , Wei Li , Junmei Wu","doi":"10.1016/j.jnucmat.2025.155723","DOIUrl":"10.1016/j.jnucmat.2025.155723","url":null,"abstract":"<div><div>Silicon Carbide (SiC) is considered a promising candidate for Accident-Tolerant Fuel cladding materials in nuclear reactors. However, existing literature often oversimplifies the heterogeneous geometric characteristics of the braided layers in SiC cladding. This paper presents a detailed modeling of the thermo-mechanical behavior of SiC cladding under light water reactor (LWR) conditions, with a focus on the braiding structure. The yarn, composed of SiC fibers and matrix, is treated as a homogenized orthogonal anisotropic material, and the braiding structure is constructed based on the parametric equations describing yarn paths. The interlayer damage between the yarns and the matrix is modeled using the cohesive zone method. The thermal-mechanical performance of the SiC<sub>f</sub>/SiC cladding during reactor startup, power operation and reactor shutdown is evaluated. The results confirm a significant increase in the tensile stress of the braided layer during reactor shutdown. Varying the anisotropic swelling of yarns only have slight effect on the cladding stress. Furthermore, the impacts of braiding patterns, braiding angles and fuel rod gap pressure are also investigated. The findings contribute to a more realistic assessment of SiC composite cladding performance, potentially informing future cladding design and fuel safety assessment.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"608 ","pages":"Article 155723"},"PeriodicalIF":2.8,"publicationDate":"2025-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143562643","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2025-03-01DOI: 10.1016/j.jnucmat.2025.155724
Hengchang Liu , Jie Shen , Jing Chen , Hongniao Chen , Yuxiang Tang
Three-point bending tests were conducted on single edge center-notched beams to study the fracture behavior of nuclear graphite IG11. Full-field information of surface deformation of the specimens was measured by digital image correlation (DIC) technique while internal micro-cracks and damage of the specimen were monitored using acoustic emission (AE) technique. The initiation and evolution of crack and fracture process zone (FPZ) were studied and the factors affecting the behavior of KR-curves were investigated. Based on the test results, fracture toughness KIc and fracture energy GFP-δ of IG11 graphite were determined and the FPZ length was quantitatively evaluated using the strain thresholds. The FPZ initiated at 55 % of the peak load and preceded the crack initiation. The evolution of the FPZ showed three phases and reached the maximum at approximately 75 % of Pc in post-peak stage with a length of 8.43 ± 0.41 mm. The obtained KR-curves showed typical three stages: an initial rise, a stable plateau and a final decrease. The effect of the FPZ length on KR-curve behavior was discussed and factors affecting the behavior of the KR-curve were analyzed, including the length of the bridging zone, the test type and the grain size.
{"title":"Fracture behavior and KR-curve characteristics of nuclear graphite IG11 under three-point bending tests","authors":"Hengchang Liu , Jie Shen , Jing Chen , Hongniao Chen , Yuxiang Tang","doi":"10.1016/j.jnucmat.2025.155724","DOIUrl":"10.1016/j.jnucmat.2025.155724","url":null,"abstract":"<div><div>Three-point bending tests were conducted on single edge center-notched beams to study the fracture behavior of nuclear graphite IG11. Full-field information of surface deformation of the specimens was measured by digital image correlation (DIC) technique while internal micro-cracks and damage of the specimen were monitored using acoustic emission (AE) technique. The initiation and evolution of crack and fracture process zone (FPZ) were studied and the factors affecting the behavior of <em>K</em><sub>R</sub>-curves were investigated. Based on the test results, fracture toughness <em>K</em><sub>Ic</sub> and fracture energy <em>G</em><sub>F</sub><em><sup>P-δ</sup></em> of IG11 graphite were determined and the FPZ length was quantitatively evaluated using the strain thresholds. The FPZ initiated at 55 % of the peak load and preceded the crack initiation. The evolution of the FPZ showed three phases and reached the maximum at approximately 75 % of P<sub>c</sub> in post-peak stage with a length of 8.43 ± 0.41 mm. The obtained <em>K</em><sub>R</sub>-curves showed typical three stages: an initial rise, a stable plateau and a final decrease. The effect of the FPZ length on <em>K</em><sub>R</sub>-curve behavior was discussed and factors affecting the behavior of the <em>K</em><sub>R</sub>-curve were analyzed, including the length of the bridging zone, the test type and the grain size.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"608 ","pages":"Article 155724"},"PeriodicalIF":2.8,"publicationDate":"2025-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143562642","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2025-03-01DOI: 10.1016/j.jnucmat.2025.155707
S. van Til , A.V. Fedorov , F. Nindiyasari , F. Charpin-Jacobs , G. Uitslag , F. Pasti , E. D'Agata , N. Chauvin
To explore fuel operational behaviour and material property evolution under Pu-burning conditions for fast reactor application, several (U,Pu)O2 MOX fuel pins with increased Pu contents (40 %HM) were irradiated in the HFR Petten in the TRABANT-2 experiment. Fuel pin number 2 (pin 2/2 in short) was designed and produced in the CAPRA programme [1], containing annular (U,Pu)O2 MOX pellets with a Pu content of 40 % (HM), that were fabricated via classic powder metallurgy, loaded into an austenitic steel cladding tube (15–15Ti). The pin was assembled and immersed in a sodium-filled experimental capsule and irradiated in the High Flux Reactor at a linear heat rate (LHR) of 450–480W/cm and with cladding temperatures not exceeding 600 °C. The irradiation was stopped after three irradiation cycles (74 days) after strong mobility of the central hole was observed in the pellets in neutron radiographs, indicating unexpected high central temperatures.
The post-irradiation neutronics analysis, using neutron fluence detectors located close to the pin confirms a maximum LHR of 447 W/cm. Asymmetric central hole growth and relocation was observed in fuel pin regions exceeding LHR of 407 W/cm.
The temperature history was reconstructed, using instrumentation in the HFR and the sample holder and Post Irradiation Examinations (PIE) on this fuel pin are carried out NRG's Hot Cell Laboratories within the European H2020 project PuMMA [2].
This paper presents a reconstruction of the irradiation history, results of a set of non-destructive examinations (NDE) and fission gas release analysis. The underlying phenomenological explanation on the observed asymmetries is presented and preliminary confirmed by a 2D thermal-mechanical model.
{"title":"First Post Irradiation Examinations on a fast reactor grade MOX fuel (U0.6,Pu0.4)O2 for Pu-burning application, irradiated in the High Flux Reactor","authors":"S. van Til , A.V. Fedorov , F. Nindiyasari , F. Charpin-Jacobs , G. Uitslag , F. Pasti , E. D'Agata , N. Chauvin","doi":"10.1016/j.jnucmat.2025.155707","DOIUrl":"10.1016/j.jnucmat.2025.155707","url":null,"abstract":"<div><div>To explore fuel operational behaviour and material property evolution under Pu-burning conditions for fast reactor application, several (U,Pu)O<sub>2</sub> MOX fuel pins with increased Pu contents (40 %HM) were irradiated in the HFR Petten in the TRABANT-2 experiment. Fuel pin number 2 (pin 2/2 in short) was designed and produced in the CAPRA programme [1], containing annular (U,Pu)O<sub>2</sub> MOX pellets with a Pu content of 40 % (HM), that were fabricated via classic powder metallurgy, loaded into an austenitic steel cladding tube (15–15Ti). The pin was assembled and immersed in a sodium-filled experimental capsule and irradiated in the High Flux Reactor at a linear heat rate (LHR) of 450–480W/cm and with cladding temperatures not exceeding 600 °C. The irradiation was stopped after three irradiation cycles (74 days) after strong mobility of the central hole was observed in the pellets in neutron radiographs, indicating unexpected high central temperatures.</div><div>The post-irradiation neutronics analysis, using neutron fluence detectors located close to the pin confirms a maximum LHR of 447 W/cm. Asymmetric central hole growth and relocation was observed in fuel pin regions exceeding LHR of 407 W/cm.</div><div>The temperature history was reconstructed, using instrumentation in the HFR and the sample holder and Post Irradiation Examinations (PIE) on this fuel pin are carried out NRG's Hot Cell Laboratories within the European H2020 project PuMMA [2].</div><div>This paper presents a reconstruction of the irradiation history, results of a set of non-destructive examinations (NDE) and fission gas release analysis. The underlying phenomenological explanation on the observed asymmetries is presented and preliminary confirmed by a 2D thermal-mechanical model.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"608 ","pages":"Article 155707"},"PeriodicalIF":2.8,"publicationDate":"2025-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143548943","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
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