Pub Date : 2025-12-24DOI: 10.1016/j.jnucmat.2025.156412
Jaewoo Park, Daniele Salvato, Adam B. Robinson, William A. Hanson, Jan-Fong Jue, Tammy L. Trowbridge, Jeffrey J. Giglio
The stability of U-Mo fuel particles embedded in an Al matrix under irradiation can be enhanced through ZrN coatings and/or heat treatment. The present study investigates the irradiation behavior of fuel plates containing U-Mo fuel particles fabricated under various heat-treatment conditions and ZrN coating thicknesses. Different fission densities were also applied to each fuel plate to evaluate the effects of these variables.
Results indicate that higher fission densities lead to more grain recrystallization and high burnup structure (HBS) development in the fuel particles. Heat treatment was found to mitigate the accumulation of fission gas bubbles in fuel particles at low fission densities by coarsening their grains. Fuel particles with ZrN coatings of a 1.2 μm thickness or above exhibited reduced formation of U-Mo/Al interaction layers, suggesting the existence of a critical ZrN coating thickness that minimizes the development of these layers.
Fission gas bubbles were predominantly observed at grain boundaries of U-Mo fuel particles irradiated at low fission densities. Subgrain boundaries, which appeared to originate from the original grain boundaries containing fission gas bubbles or HBSs, were also observed, indicating the early stage of HBS propagation in the fuel particles.
{"title":"Effects of ZrN coating and heat treatment on U-Mo dispersion fuel systems under irradiation","authors":"Jaewoo Park, Daniele Salvato, Adam B. Robinson, William A. Hanson, Jan-Fong Jue, Tammy L. Trowbridge, Jeffrey J. Giglio","doi":"10.1016/j.jnucmat.2025.156412","DOIUrl":"10.1016/j.jnucmat.2025.156412","url":null,"abstract":"<div><div>The stability of U-Mo fuel particles embedded in an Al matrix under irradiation can be enhanced through ZrN coatings and/or heat treatment. The present study investigates the irradiation behavior of fuel plates containing U-Mo fuel particles fabricated under various heat-treatment conditions and ZrN coating thicknesses. Different fission densities were also applied to each fuel plate to evaluate the effects of these variables.</div><div>Results indicate that higher fission densities lead to more grain recrystallization and high burnup structure (HBS) development in the fuel particles. Heat treatment was found to mitigate the accumulation of fission gas bubbles in fuel particles at low fission densities by coarsening their grains. Fuel particles with ZrN coatings of a 1.2 μm thickness or above exhibited reduced formation of U-Mo/Al interaction layers, suggesting the existence of a critical ZrN coating thickness that minimizes the development of these layers.</div><div>Fission gas bubbles were predominantly observed at grain boundaries of U-Mo fuel particles irradiated at low fission densities. Subgrain boundaries, which appeared to originate from the original grain boundaries containing fission gas bubbles or HBSs, were also observed, indicating the early stage of HBS propagation in the fuel particles.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"622 ","pages":"Article 156412"},"PeriodicalIF":3.2,"publicationDate":"2025-12-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145922653","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 : 2025-12-23DOI: 10.1016/j.jnucmat.2025.156404
Yuhan Li , Chao Jiang , Tiankai Yao , Haiyan Wang , Jian Gan
Advanced fuel cladding is critical for fast reactors, offering sufficient thermal conductivity, mechanical and dimensional stability and radiation tolerance of the cladding base material. Additionally, it must provide corrosion resistance and high temperature coolant compatibility on the cladding outer surface, as well as chemical stability on the cladding inner wall against fuel cladding chemical interaction (FCCI). TiN ceramic coating has been considered an effective diffusion barrier for inner and outer cladding-walls for enhanced performance. The TiN-metal interface microstructure and chemistry play a critical role in coating bond strength and integrity under harsh conditions. High-resolution transmission electron microscopy characterization of ceramic-metal interface at atomic resolution in unirradiated, irradiated and thermal cycled conditions were performed. The interface remained intact after irradiation up to 200 dpa or thermal cycling five times up to 550 °C. This work discusses the potential impact of these results on coating performance and design for advanced claddings.
{"title":"High-resolution characterization of ceramic-metal interface of TiN coating on ferritic-steels for nuclear application","authors":"Yuhan Li , Chao Jiang , Tiankai Yao , Haiyan Wang , Jian Gan","doi":"10.1016/j.jnucmat.2025.156404","DOIUrl":"10.1016/j.jnucmat.2025.156404","url":null,"abstract":"<div><div>Advanced fuel cladding is critical for fast reactors<strong>,</strong> offering sufficient thermal conductivity, mechanical and dimensional stability and radiation tolerance of the cladding base material<strong>.</strong> Additionally<strong>,</strong> it must provide corrosion resistance and high temperature coolant compatibility on the cladding outer surface, as well as chemical stability on the cladding inner wall against fuel cladding chemical interaction (FCCI). TiN ceramic coating has been considered an effective diffusion barrier for inner and outer cladding-walls for enhanced performance. The TiN-metal interface microstructure and chemistry play a critical role in coating bond strength and integrity under harsh conditions. High-resolution transmission electron microscopy characterization of ceramic-metal interface at atomic resolution in unirradiated, irradiated and thermal cycled conditions were performed. The interface remained intact after irradiation up to 200 dpa or thermal cycling five times up to 550 °C. This work discusses the potential impact of these results on coating performance and design for advanced claddings.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"621 ","pages":"Article 156404"},"PeriodicalIF":3.2,"publicationDate":"2025-12-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145838044","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 : 2025-12-23DOI: 10.1016/j.jnucmat.2025.156411
Grant W. Helmreich , John D. Hunn , Fred C. Montgomery , Darren J. Skitt , Tamara J. Keever , Benjamin D. Roach , Kayron T. Rogers , James W. Sterbenz
Established methods for the determination of burnup in nuclear fuels commonly rely on the measurement of 148Nd in the spent fuel in concert with known fission product yields to determine the atom percent of fissions in the fuel. This isotope of neodymium is used for various reasons, including chemical and radioactive stability, ease of measurement, and low rates of formation and destruction due to neutron flux apart from fission. However, careful calculation of effective cumulative fission yields and correction factors for (n,γ) capture reactions allows for additional stable and long-lived isotopes of neodymium to be used to provide additional independent measurements of burnup, reducing statistical uncertainty. This method was developed and successfully applied to measure the burnup of compacts from the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. The mean burnup measured using 143Nd, 145Nd, 146Nd, 148Nd, and 150Nd was statistically observed to be the same as that measured using 148Nd alone, but the statistical uncertainty in the measurement was reduced by a factor of 2, providing a tighter confidence interval in the final results.
{"title":"Expanded method for the determination of burnup in nuclear fuels using multiple neodymium isotopes","authors":"Grant W. Helmreich , John D. Hunn , Fred C. Montgomery , Darren J. Skitt , Tamara J. Keever , Benjamin D. Roach , Kayron T. Rogers , James W. Sterbenz","doi":"10.1016/j.jnucmat.2025.156411","DOIUrl":"10.1016/j.jnucmat.2025.156411","url":null,"abstract":"<div><div>Established methods for the determination of burnup in nuclear fuels commonly rely on the measurement of <sup>148</sup>Nd in the spent fuel in concert with known fission product yields to determine the atom percent of fissions in the fuel. This isotope of neodymium is used for various reasons, including chemical and radioactive stability, ease of measurement, and low rates of formation and destruction due to neutron flux apart from fission. However, careful calculation of effective cumulative fission yields and correction factors for (n,γ) capture reactions allows for additional stable and long-lived isotopes of neodymium to be used to provide additional independent measurements of burnup, reducing statistical uncertainty. This method was developed and successfully applied to measure the burnup of compacts from the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. The mean burnup measured using <sup>143</sup>Nd, <sup>145</sup>Nd, <sup>146</sup>Nd, <sup>148</sup>Nd, and <sup>150</sup>Nd was statistically observed to be the same as that measured using <sup>148</sup>Nd alone, but the statistical uncertainty in the measurement was reduced by a factor of 2, providing a tighter confidence interval in the final results.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"622 ","pages":"Article 156411"},"PeriodicalIF":3.2,"publicationDate":"2025-12-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145882263","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}
This work presents and validates a novel approach for modelling non-inert fission products behaviour during a power transient. The proposed model, MARGARET Active Fission Products, leverages all the microstructural quantities computed by MARGARET-a code originally developed to evaluate the release of inert fission products under normal and incidental loading sequences-to simulate the production, decay, and transport of non-inert fission products. The MARGARET Active Fission Products model is capable of handling 75 different isotopes, ranging from short-lived to long-lived species. The tendency of non-inert fission products to form different chemical compounds in different phases within the fuel compared to inert ones (always gaseous at equilibrium), requires the incorporation of thermochemistry in the calculations, since it directly impacts their release kinetics from the fuel. The thermochemical behaviour of fission products is included by performing calculations with the OpenCalphad thermochemical solver and the TAF-ID database. In this coupled thermochemistry-fission gas release approach, thermochemistry leads to the assessment of the quantity of non-inert fission products in the gas phase, that will percolate towards the free volume and, consequently, be released in the fuel rod.
To validate the model, a power transient simulation was performed, replicating an experiment conducted on a refabricated fuel rod in the OSIRIS experimental reactor. The simulation results are compared with experimental data, including the total xenon release (measured by puncturing), the release of 137Cs, 131I, and 132Te at each interpellet (obtained from gamma spectrometry), and the radial distribution of xenon in the pellet, before and after the power ramp (analysed by Secondary-Ion Mass Spectrometry measurements). Furthermore, a detailed discussion on the thermochemical results is provided.
{"title":"Mechanistic modelling of chemically reactive fission product release during power ramps","authors":"Giacomo Petrosillo , Jérôme Sercombe , Laurence Noirot , Clément Introïni , Bertrand Baurens , Lelio Luzzi , Yves Pontillon","doi":"10.1016/j.jnucmat.2025.156400","DOIUrl":"10.1016/j.jnucmat.2025.156400","url":null,"abstract":"<div><div>This work presents and validates a novel approach for modelling non-inert fission products behaviour during a power transient. The proposed model, MARGARET Active Fission Products, leverages all the microstructural quantities computed by MARGARET-a code originally developed to evaluate the release of inert fission products under normal and incidental loading sequences-to simulate the production, decay, and transport of non-inert fission products. The MARGARET Active Fission Products model is capable of handling 75 different isotopes, ranging from short-lived to long-lived species. The tendency of non-inert fission products to form different chemical compounds in different phases within the fuel compared to inert ones (always gaseous at equilibrium), requires the incorporation of thermochemistry in the calculations, since it directly impacts their release kinetics from the fuel. The thermochemical behaviour of fission products is included by performing calculations with the OpenCalphad thermochemical solver and the TAF-ID database. In this coupled thermochemistry-fission gas release approach, thermochemistry leads to the assessment of the quantity of non-inert fission products in the gas phase, that will percolate towards the free volume and, consequently, be released in the fuel rod.</div><div>To validate the model, a power transient simulation was performed, replicating an experiment conducted on a refabricated fuel rod in the OSIRIS experimental reactor. The simulation results are compared with experimental data, including the total xenon release (measured by puncturing), the release of <sup>137</sup>Cs, <sup>131</sup>I, and <sup>132</sup>Te at each interpellet (obtained from gamma spectrometry), and the radial distribution of xenon in the pellet, before and after the power ramp (analysed by Secondary-Ion Mass Spectrometry measurements). Furthermore, a detailed discussion on the thermochemical results is provided.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"622 ","pages":"Article 156400"},"PeriodicalIF":3.2,"publicationDate":"2025-12-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145882268","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 : 2025-12-20DOI: 10.1016/j.jnucmat.2025.156399
Joel L. Abraham , Pranesh Dayal , Rifat Farzana , Ghazaleh Bahmanrokh , Charles C. Sorrell , Pramod Koshy , Daniel J. Gregg
Zirconolite is a candidate wasteform for actinide immobilisation. The addition of glass to form a glass-ceramic (GC) is also under consideration as GC materials provide flexibility to immobilise heterogeneous actinide wastes and simplify processing requirements. However, a major challenge in the design of zirconolite GCs is control of the phase assemblage to minimise unwanted phase formation, particularly at high glass contents where zirconolite can be destabilised in the glass melt during consolidation. In the current research, an optimal glass composition was developed to minimise unwanted secondary phases. Initially, GCs targeting zirconolite (CaZrTi2O7) with varying amounts (0–100 wt%) of glass addition (NaAl0.5B0.5Si2O6) were fabricated using a pre-synthesis route. X-ray diffraction (XRD) analysis of these baseline formulations showed that undesired phases (e.g., zircon) became more apparent at higher glass contents (e.g., 75 wt%). Following this, the additions of Al2O3, CaO, and TiO2 to the glass composition minimised unwanted phase formation in the GCs, including those formulations with high glass contents. The optimal glass composition was determined to be NaAl1.5Ca0.7Ti0.2B0.5Si2O8.6. Ce-bearing zirconolite GCs (Ca0.8Ce0.2ZrTi1.6Al0.4O7; Ce as actinide surrogate) with varying amounts (0–100 vol%) of the tailored glass design (NaAl1.5Ca0.7Ti0.2B0.5Si2O8.6) were then fabricated using an in-situ crystallisation route. X-ray diffraction (XRD) and scanning electron microscopy (SEM) analyses showed that near phase-pure microstructures were achieved across all glass contents. Furthermore, the addition of glass lowered the sintering temperature (1320 °C to 1270 °C) needed to immobilise CeO2 in zirconolite.
{"title":"Optimising glass-ceramic compositions for zirconolite-based actinide immobilisation","authors":"Joel L. Abraham , Pranesh Dayal , Rifat Farzana , Ghazaleh Bahmanrokh , Charles C. Sorrell , Pramod Koshy , Daniel J. Gregg","doi":"10.1016/j.jnucmat.2025.156399","DOIUrl":"10.1016/j.jnucmat.2025.156399","url":null,"abstract":"<div><div>Zirconolite is a candidate wasteform for actinide immobilisation. The addition of glass to form a glass-ceramic (GC) is also under consideration as GC materials provide flexibility to immobilise heterogeneous actinide wastes and simplify processing requirements. However, a major challenge in the design of zirconolite GCs is control of the phase assemblage to minimise unwanted phase formation, particularly at high glass contents where zirconolite can be destabilised in the glass melt during consolidation. In the current research, an optimal glass composition was developed to minimise unwanted secondary phases. Initially, GCs targeting zirconolite (CaZrTi<sub>2</sub>O<sub>7</sub>) with varying amounts (0–100 wt%) of glass addition (NaAl<sub>0.5</sub>B<sub>0.5</sub>Si<sub>2</sub>O<sub>6</sub>) were fabricated using a pre-synthesis route. X-ray diffraction (XRD) analysis of these baseline formulations showed that undesired phases (e.g., zircon) became more apparent at higher glass contents (e.g., 75 wt%). Following this, the additions of Al<sub>2</sub>O<sub>3</sub>, CaO, and TiO<sub>2</sub> to the glass composition minimised unwanted phase formation in the GCs, including those formulations with high glass contents. The optimal glass composition was determined to be NaAl<sub>1.5</sub>Ca<sub>0.7</sub>Ti<sub>0.2</sub>B<sub>0.5</sub>Si<sub>2</sub>O<sub>8.6</sub>. Ce-bearing zirconolite GCs (Ca<sub>0.8</sub>Ce<sub>0.2</sub>ZrTi<sub>1.6</sub>Al<sub>0.4</sub>O<sub>7</sub>; Ce as actinide surrogate) with varying amounts (0–100 vol%) of the tailored glass design (NaAl<sub>1.5</sub>Ca<sub>0.7</sub>Ti<sub>0.2</sub>B<sub>0.5</sub>Si<sub>2</sub>O<sub>8.6</sub>) were then fabricated using an <em>in-situ</em> crystallisation route. X-ray diffraction (XRD) and scanning electron microscopy (SEM) analyses showed that near phase-pure microstructures were achieved across all glass contents. Furthermore, the addition of glass lowered the sintering temperature (1320 °C to 1270 °C) needed to immobilise CeO<sub>2</sub> in zirconolite.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"621 ","pages":"Article 156399"},"PeriodicalIF":3.2,"publicationDate":"2025-12-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145838103","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 : 2025-12-20DOI: 10.1016/j.jnucmat.2025.156401
Wentao Qin , Weifeng Liu , Chunjie Niu , Weiyuan Ni , Dongping Liu
Based on our recent study (Nucl. Fusion 64 (2024) 056,039), an improved model is employed to investigate the impact of helium bubbles in tungsten (W) on hydrogen isotope retention. Simulations were conducted to assess the changes in helium bubble size and density in W under helium pre-irradiation at the W temperature ranging from 400 K to 2300 K. Subsequently, the effect of these helium bubbles on hydrogen isotope retention was examined through simulated hydrogen isotope irradiation at 400 K. As the W surface temperature increases during helium pre-irradiation, the helium bubbles exhibit size growth with concomitant density reduction. Correspondingly, the hydrogen isotope retention increases and then decreases with the rise in W surface temperature during helium pre-irradiation. When the hydrogen dose is further increased, the hydrogen retention gradually increases. The application of rate theory models in our study helps in effectively simulating the long-term evolution and large-scale behavior of hydrogen and helium interactions in W under fusion-relevant conditions, providing a more comprehensive understanding. The research demonstrates that the temperature of helium irradiation significantly influences the characteristics of helium bubbles in W, which in turn affects the retention of hydrogen isotopes. This work contributes to the understanding and optimization of fusion reactor wall materials, promoting improved performance and reliability.
{"title":"Numerical simulation study on the impact of helium bubbles on hydrogen isotope retention behavior in tungsten","authors":"Wentao Qin , Weifeng Liu , Chunjie Niu , Weiyuan Ni , Dongping Liu","doi":"10.1016/j.jnucmat.2025.156401","DOIUrl":"10.1016/j.jnucmat.2025.156401","url":null,"abstract":"<div><div>Based on our recent study (Nucl. Fusion 64 (2024) 056,039), an improved model is employed to investigate the impact of helium bubbles in tungsten (W) on hydrogen isotope retention. Simulations were conducted to assess the changes in helium bubble size and density in W under helium pre-irradiation at the W temperature ranging from 400 K to 2300 K. Subsequently, the effect of these helium bubbles on hydrogen isotope retention was examined through simulated hydrogen isotope irradiation at 400 K. As the W surface temperature increases during helium pre-irradiation, the helium bubbles exhibit size growth with concomitant density reduction. Correspondingly, the hydrogen isotope retention increases and then decreases with the rise in W surface temperature during helium pre-irradiation. When the hydrogen dose is further increased, the hydrogen retention gradually increases. The application of rate theory models in our study helps in effectively simulating the long-term evolution and large-scale behavior of hydrogen and helium interactions in W under fusion-relevant conditions, providing a more comprehensive understanding. The research demonstrates that the temperature of helium irradiation significantly influences the characteristics of helium bubbles in W, which in turn affects the retention of hydrogen isotopes. This work contributes to the understanding and optimization of fusion reactor wall materials, promoting improved performance and reliability.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"622 ","pages":"Article 156401"},"PeriodicalIF":3.2,"publicationDate":"2025-12-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145882265","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 : 2025-12-20DOI: 10.1016/j.jnucmat.2025.156397
Miaomiao Hu , Xinmei Yang , Huajian Liu , Xingtai Zhou
The corrosion mechanisms of nickel-based alloy (UNS N10003) and stainless steel (316H SS) in molten LiF-NaF-KF (FLiNaK) salt at 700 °C were investigated. Results indicate that UNS N10003 exhibits better resistance to intergranular corrosion than 316H SS in the molten FLiNaK salt with and without Te. In the same batch of molten FLiNaK salt without Te, for UNS N10003 alloy, the reactions between its alloying elements and the impurities in the salt caused the uniform corrosion; while the reactions led to an intergranular corrosion for 316H SS. The presence of Te in molten FLiNaK salt induced the intergranular corrosion for UNS N10003 alloy. The coexistence of Te (1 wt% Te) and the impurities in the molten FLiNaK salt aggravated the corrosion of UNS N10003 alloy and 316H SS by ∼12 times. The corrosion depth of 316H SS (∼370 μm for 400 h) is larger than that of UNS N10003 (∼90 μm for 400 h) in the molten FLiNaK salt with Te (1 wt%). The severer intergranular corrosion is mainly attributed to the reactions involving Te, the impurities in salt, the alloying elements, and the precipitates at grain boundaries.
{"title":"Combined effect of Te and impurities in molten LiF-NaF-KF salt on the corrosion of 316H SS and UNS N10003 alloy","authors":"Miaomiao Hu , Xinmei Yang , Huajian Liu , Xingtai Zhou","doi":"10.1016/j.jnucmat.2025.156397","DOIUrl":"10.1016/j.jnucmat.2025.156397","url":null,"abstract":"<div><div>The corrosion mechanisms of nickel-based alloy (UNS N10003) and stainless steel (316H SS) in molten LiF-NaF-KF (FLiNaK) salt at 700 °C were investigated. Results indicate that UNS N10003 exhibits better resistance to intergranular corrosion than 316H SS in the molten FLiNaK salt with and without Te. In the same batch of molten FLiNaK salt without Te, for UNS N10003 alloy, the reactions between its alloying elements and the impurities in the salt caused the uniform corrosion; while the reactions led to an intergranular corrosion for 316H SS. The presence of Te in molten FLiNaK salt induced the intergranular corrosion for UNS N10003 alloy. The coexistence of Te (1 wt% Te) and the impurities in the molten FLiNaK salt aggravated the corrosion of UNS N10003 alloy and 316H SS by ∼12 times. The corrosion depth of 316H SS (∼370 μm for 400 h) is larger than that of UNS N10003 (∼90 μm for 400 h) in the molten FLiNaK salt with Te (1 wt%). The severer intergranular corrosion is mainly attributed to the reactions involving Te, the impurities in salt, the alloying elements, and the precipitates at grain boundaries.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"621 ","pages":"Article 156397"},"PeriodicalIF":3.2,"publicationDate":"2025-12-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145838099","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 : 2025-12-19DOI: 10.1016/j.jnucmat.2025.156393
Lu Liang , Lingyan Xu , Qinzeng Hu , Yingming Wang , Zhentao Qin , Yanyan Lei , Wei Zheng , Shuai Song , Chaopeng Mi , Roman Lanovsky , Wanqi Jie
Understanding the irradiation damage of cadmium zinc telluride (CdZnTe, CZT) crystals and its effect on photoelectric properties is crucial for their reliable use in radiation detection. This study examines the combined influence of electronic (Se) and nuclear (Sn) energy loss on the microstructure, current transport, and carrier characteristics of CZT using 516 MeV and 1.5 MeV Xe ion irradiations. Results indicate that the synergistic Se/Sn effect critically influences defect evolution pathways, leading to divergent microstructures. High Se favors dislocation loops through thermal-spike-enhanced kinetics, whereas high Sn, by exacerbating lattice damage, promotes stacking faults and the evolution of loops into large-scale dislocation lines. Defect levels are deeper and the concentration of defects is larger after 516 MeV irradiation than after 1.5 MeV. Leakage current mechanisms are dominated by Schottky emission (SE) combined with the Poole-Frenkel (PF) effect for 516 MeV Xe ions, and by Fowler-Nordheim (F-N) tunneling coupled with PF effect for 1.5 MeV Xe ions. Carrier transport and γ-ray detection performance degrade more severely under 516 MeV irradiation, likely because of its broader damage layer and deeper defect levels. These findings provide a theoretical basis for understanding radiation damage mechanisms and performance recovery in CZT, offering valuable insights for radiation protection in spaceborne equipment.
{"title":"Synergistic roles of electronic and nuclear energy deposition: From defect generation to performance degradation in heavy-ion-irradiated CdZnTe crystals","authors":"Lu Liang , Lingyan Xu , Qinzeng Hu , Yingming Wang , Zhentao Qin , Yanyan Lei , Wei Zheng , Shuai Song , Chaopeng Mi , Roman Lanovsky , Wanqi Jie","doi":"10.1016/j.jnucmat.2025.156393","DOIUrl":"10.1016/j.jnucmat.2025.156393","url":null,"abstract":"<div><div>Understanding the irradiation damage of cadmium zinc telluride (CdZnTe, CZT) crystals and its effect on photoelectric properties is crucial for their reliable use in radiation detection. This study examines the combined influence of electronic (S<sub>e</sub>) and nuclear (S<sub>n</sub>) energy loss on the microstructure, current transport, and carrier characteristics of CZT using 516 MeV and 1.5 MeV Xe ion irradiations. Results indicate that the synergistic S<sub>e</sub>/S<sub>n</sub> effect critically influences defect evolution pathways, leading to divergent microstructures. High S<sub>e</sub> favors dislocation loops through thermal-spike-enhanced kinetics, whereas high S<sub>n</sub>, by exacerbating lattice damage, promotes stacking faults and the evolution of loops into large-scale dislocation lines. Defect levels are deeper and the concentration of defects is larger after 516 MeV irradiation than after 1.5 MeV. Leakage current mechanisms are dominated by Schottky emission (SE) combined with the Poole-Frenkel (PF) effect for 516 MeV Xe ions, and by Fowler-Nordheim (F-N) tunneling coupled with PF effect for 1.5 MeV Xe ions. Carrier transport and γ-ray detection performance degrade more severely under 516 MeV irradiation, likely because of its broader damage layer and deeper defect levels. These findings provide a theoretical basis for understanding radiation damage mechanisms and performance recovery in CZT, offering valuable insights for radiation protection in spaceborne equipment.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"621 ","pages":"Article 156393"},"PeriodicalIF":3.2,"publicationDate":"2025-12-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145838101","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 : 2025-12-19DOI: 10.1016/j.jnucmat.2025.156395
Wenjie Zhang, Zhenyu Jiang, Honghui Zhang, Tianyi Song, Yong Liu, Yao Zhang, Ze Li, Yubin Pan, Kaigui Zhu
One of the challenges in fusion reactor research is the development of plasma-facing materials that can endure extreme environments. Although tungsten (W) is regarded as the most promising plasma-facing material, its performance inevitably degrades under fusion reactor conditions. When exposed to low-energy, high-flux hydrogen isotope plasma, W will experience a risk of structural damage. Meanwhile, neutron irradiation will generate transmutation products of W, which can alter the mechanical performance of W alloys. In this work, W-rhenium (Re) films with homogeneous deuterium (D) distribution are manufactured by magnetron sputtering to investigate the effects of Re and D on the mechanical performance of W. The damage gradient effect of the prepared samples is effectively mitigated, which has significant advantages compared to samples prepared by ion implantation or plasma exposure. The results indicate that sample hardness would increase with the higher D content, and the hardness-depth relationship follows the Nix-Gao model. At the same D concentration, increasing Re content leads to a reduction in hardness, which can be attributed to enhanced dislocation mobility. Meanwhile, the dislocation density in the sample has a positive relation with D content, which is consistent with the calculation results based on dispersed barrier-hardening model. The increased Re content further reduces D retention in W-Re films by decreasing the concentration of vacancy defects. This study elucidates the effects of Re and D on the mechanical properties of W under fusion reactor conditions.
{"title":"Effect of D on the microstructure evolution and hardness of W-Re films deposited by magnetron sputtering","authors":"Wenjie Zhang, Zhenyu Jiang, Honghui Zhang, Tianyi Song, Yong Liu, Yao Zhang, Ze Li, Yubin Pan, Kaigui Zhu","doi":"10.1016/j.jnucmat.2025.156395","DOIUrl":"10.1016/j.jnucmat.2025.156395","url":null,"abstract":"<div><div>One of the challenges in fusion reactor research is the development of plasma-facing materials that can endure extreme environments. Although tungsten (W) is regarded as the most promising plasma-facing material, its performance inevitably degrades under fusion reactor conditions. When exposed to low-energy, high-flux hydrogen isotope plasma, W will experience a risk of structural damage. Meanwhile, neutron irradiation will generate transmutation products of W, which can alter the mechanical performance of W alloys. In this work, W-rhenium (<em>Re</em>) films with homogeneous deuterium (D) distribution are manufactured by magnetron sputtering to investigate the effects of <em>Re</em> and D on the mechanical performance of W. The damage gradient effect of the prepared samples is effectively mitigated, which has significant advantages compared to samples prepared by ion implantation or plasma exposure. The results indicate that sample hardness would increase with the higher D content, and the hardness-depth relationship follows the Nix-Gao model. At the same D concentration, increasing <em>Re</em> content leads to a reduction in hardness, which can be attributed to enhanced dislocation mobility. Meanwhile, the dislocation density in the sample has a positive relation with D content, which is consistent with the calculation results based on dispersed barrier-hardening model. The increased <em>Re</em> content further reduces D retention in W-<em>Re</em> films by decreasing the concentration of vacancy defects. This study elucidates the effects of <em>Re</em> and D on the mechanical properties of W under fusion reactor conditions.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"621 ","pages":"Article 156395"},"PeriodicalIF":3.2,"publicationDate":"2025-12-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145838100","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 : 2025-12-19DOI: 10.1016/j.jnucmat.2025.156398
Nigel Lucas , Richard Woods , Simon Crombleholme , Hruthik Vandanapu , Connor Beer , Jeremie Sobel , Thomas Steenberg , Maulik K. Patel
This study investigated the influence of impurities on the corrosion behaviour of 316 L stainless steel in molten FLiNaK at 600 °C and LiThF at 700 °C for up to 3000 h. Coupons exposed to untreated FLiNaK containing moisture and oxides exhibited significantly higher mass loss compared to those exposed to purified salt. This coincides with the depletion of chromium (Cr) and iron (Fe) from the steel, as confirmed by ICP-OES analysis of post-experiment salts. SEM analysis identified intergranular corrosion as the primary attack mode in untreated FLiNaK, with corrosion depths up to 112 μm. Conversely, coupons exposed to purified FLiNaK displayed excellent corrosion resistance, despite trace levels of metallic impurities detected in the salt. These impurities are believed to originate from the purification process itself and may contribute to a small degree of observed corrosion. The typical mode of grain boundary Cr dissolution was observed in coupons tested in untreated FLiNaK salts, while this was absent in coupons tested in purified FLiNaK salts. Interestingly, XRD analysis identified chromium carbide phases on coupons from purified FLiNaK tests. Similar trends were observed in LiThF salt, where untreated salt resulted in severe corrosion compared to purified salt. Overall, this study highlights the critical role of salt purity in minimizing corrosion of 316 L stainless steel in molten fluoride salts. The data from the present study indicates that a purified molten FLiNaK at 600 °C can be a long-term non-corrosive environment for 316 L stainless steel.
{"title":"Effect of salt purity on the corrosion of 316 L SS: Long-term studies in molten FLiNaK and ThF4 - LiF","authors":"Nigel Lucas , Richard Woods , Simon Crombleholme , Hruthik Vandanapu , Connor Beer , Jeremie Sobel , Thomas Steenberg , Maulik K. Patel","doi":"10.1016/j.jnucmat.2025.156398","DOIUrl":"10.1016/j.jnucmat.2025.156398","url":null,"abstract":"<div><div>This study investigated the influence of impurities on the corrosion behaviour of 316 L stainless steel in molten FLiNaK at 600 °C and LiThF at 700 °C for up to 3000 h. Coupons exposed to untreated FLiNaK containing moisture and oxides exhibited significantly higher mass loss compared to those exposed to purified salt. This coincides with the depletion of chromium (Cr) and iron (Fe) from the steel, as confirmed by ICP-OES analysis of post-experiment salts. SEM analysis identified intergranular corrosion as the primary attack mode in untreated FLiNaK, with corrosion depths up to 112 μm. Conversely, coupons exposed to purified FLiNaK displayed excellent corrosion resistance, despite trace levels of metallic impurities detected in the salt. These impurities are believed to originate from the purification process itself and may contribute to a small degree of observed corrosion. The typical mode of grain boundary Cr dissolution was observed in coupons tested in untreated FLiNaK salts, while this was absent in coupons tested in purified FLiNaK salts. Interestingly, XRD analysis identified chromium carbide phases on coupons from purified FLiNaK tests. Similar trends were observed in LiThF salt, where untreated salt resulted in severe corrosion compared to purified salt. Overall, this study highlights the critical role of salt purity in minimizing corrosion of 316 L stainless steel in molten fluoride salts. The data from the present study indicates that a purified molten FLiNaK at 600 °C can be a long-term non-corrosive environment for 316 L stainless steel.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"622 ","pages":"Article 156398"},"PeriodicalIF":3.2,"publicationDate":"2025-12-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145882272","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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