Pub Date : 2025-03-30DOI: 10.1016/j.jnucmat.2025.155798
Shaobo Yang , Chenxi Liang , Jiali Li , Yujie Ma , Sijie Kou , Juanli Deng , Bo Chen , Shangwu Fan
The study investigated the corrosion behavior and mechanical performance of SiC/SiC composite joints with Y2O3-Al2O3-SiO2 (YAS) interlayers under high-temperature steam environments at 1200 °C. Under low-flow conditions, partial disruption of Si-O and Al-O bonds in the YAS glass network reduced crosslinking, forming an aluminosilicate protective layer that inhibited further corrosion. Prolonged exposure led to Y3+ migration and accumulation, resulting in Y2Si2O7 precipitation and growth. High-flow conditions caused a thinner glass layer, continuous longitudinal cracks, and more severe erosion and dissolution of the YAS glass due to higher steam velocity. Despite these degradations, the joints exhibited satisfactory performance, maintaining shear strengths of about 40 ± 2 MPa after 15 h of low-flow exposure and about 36 ± 5 MPa after 5 h of high-flow exposure. These findings demonstrate that YAS interlayers provide excellent corrosion resistance and mechanical stability as a sealant for nuclear-grade SiC/SiC.
{"title":"Corrosion behavior and mechanical properties of SiC/SiC composite joints with Y2O3-Al2O3-SiO2 interlayer under high-temperature steam environments at 1200 °C","authors":"Shaobo Yang , Chenxi Liang , Jiali Li , Yujie Ma , Sijie Kou , Juanli Deng , Bo Chen , Shangwu Fan","doi":"10.1016/j.jnucmat.2025.155798","DOIUrl":"10.1016/j.jnucmat.2025.155798","url":null,"abstract":"<div><div>The study investigated the corrosion behavior and mechanical performance of SiC/SiC composite joints with Y<sub>2</sub>O<sub>3</sub>-Al<sub>2</sub>O<sub>3</sub>-SiO<sub>2</sub> (YAS) interlayers under high-temperature steam environments at 1200 °C. Under low-flow conditions, partial disruption of Si-O and Al-O bonds in the YAS glass network reduced crosslinking, forming an aluminosilicate protective layer that inhibited further corrosion. Prolonged exposure led to Y<sup>3+</sup> migration and accumulation, resulting in Y<sub>2</sub>Si<sub>2</sub>O<sub>7</sub> precipitation and growth. High-flow conditions caused a thinner glass layer, continuous longitudinal cracks, and more severe erosion and dissolution of the YAS glass due to higher steam velocity. Despite these degradations, the joints exhibited satisfactory performance, maintaining shear strengths of about 40 ± 2 MPa after 15 h of low-flow exposure and about 36 ± 5 MPa after 5 h of high-flow exposure. These findings demonstrate that YAS interlayers provide excellent corrosion resistance and mechanical stability as a sealant for nuclear-grade SiC/SiC.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"610 ","pages":"Article 155798"},"PeriodicalIF":2.8,"publicationDate":"2025-03-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143768652","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}
Microstructures, mechanical properties and irradiation hardening of novel multi-element iron-based alloys (Fe-(10 and 20)Mn-15Cr-2.0Al-0.7V-0.5C (at %)) were investigated. The alloys do not contain high activation elements, such as, Co, Ni and Mo. The alloy samples were hot-rolled at 1323 K and air-cooled, followed by heat treatment at 1073 K for 0.5 h and quenching in to water. After the heat-treatment, the Fe-10Mn-15Cr-2.0Al-0.7V-0.5C (10Mn) sample consisted mainly of body-centered cubic (BCC) structure with two distinct microstructures, i.e., fine lath-martensite-like structures and recrystallized grains. Meanwhile, the Fe-20Mn-15Cr-2.0Al-0.7V-0.5C (20Mn) sample were a mixture of fine lath-martensite like BCC phase and face-centered cubic (FCC) phases. The 10Mn sample exhibits very high tensile strength of 960 MPa but low elongation, while the 20Mn sample exhibits lower tensile strength of 620 MPa but much improved elongation over 60 %. The samples were simultaneously triple-irradiated with 10.5 MeV Fe3+ ions, 1.05 MeV He+ ions and 0.38 MeV H+ ions to a depth of 1 μm from the sample surface. The irradiation hardening in average was only about 1.5 GPa in the alloys irradiated with 10.5 MeV Fe3+ ions up to 30 dpa at 573 K at the damage peak, measured by nano-indentation. The irradiation hardening resistance of the alloys was better than that of other fusion structural materials and fission reactor pressure vessel steels. Combined analysis with electron-backscattered diffraction and nanoindentation revealed that the irradiation hardening is less significant in lath BCC phase than in recrystallized BCC (10Mn) and in FCC (20Mn). These results suggest that the alloys with good combination of irradiation resistance and mechanical properties can be developed by further tailoring the phase stability of the alloys and combining the high-entropy effects, aiming for the application for components in nuclear reactors, fusion reactors and high-power large accelerator facilities.
{"title":"Development of novel multi-element low-activation Fe-based alloys for nuclear and fusion reactor applications","authors":"Kazuyuki Furuya , Koichi Tsuchiya , Eiichi Wakai , Elango Chandiran , Bikash Tripathy , Masami Ando , Takaharu Kamada , Hiroyuki Noto","doi":"10.1016/j.jnucmat.2025.155772","DOIUrl":"10.1016/j.jnucmat.2025.155772","url":null,"abstract":"<div><div>Microstructures, mechanical properties and irradiation hardening of novel multi-element iron-based alloys (Fe-(10 and 20)Mn-15Cr-2.0Al-0.7V-0.5C (at %)) were investigated. The alloys do not contain high activation elements, such as, Co, Ni and Mo. The alloy samples were hot-rolled at 1323 K and air-cooled, followed by heat treatment at 1073 K for 0.5 h and quenching in to water. After the heat-treatment, the Fe-10Mn-15Cr-2.0Al-0.7V-0.5C (10Mn) sample consisted mainly of body-centered cubic (BCC) structure with two distinct microstructures, i.e., fine lath-martensite-like structures and recrystallized grains. Meanwhile, the Fe-20Mn-15Cr-2.0Al-0.7V-0.5C (20Mn) sample were a mixture of fine lath-martensite like BCC phase and face-centered cubic (FCC) phases. The 10Mn sample exhibits very high tensile strength of 960 MPa but low elongation, while the 20Mn sample exhibits lower tensile strength of 620 MPa but much improved elongation over 60 %. The samples were simultaneously triple-irradiated with 10.5 MeV Fe<sup>3+</sup> ions, 1.05 MeV He<sup>+</sup> ions and 0.38 MeV H<sup>+</sup> ions to a depth of 1 μm from the sample surface. The irradiation hardening in average was only about 1.5 GPa in the alloys irradiated with 10.5 MeV Fe<sup>3+</sup> ions up to 30 dpa at 573 K at the damage peak, measured by nano-indentation. The irradiation hardening resistance of the alloys was better than that of other fusion structural materials and fission reactor pressure vessel steels. Combined analysis with electron-backscattered diffraction and nanoindentation revealed that the irradiation hardening is less significant in lath BCC phase than in recrystallized BCC (10Mn) and in FCC (20Mn). These results suggest that the alloys with good combination of irradiation resistance and mechanical properties can be developed by further tailoring the phase stability of the alloys and combining the high-entropy effects, aiming for the application for components in nuclear reactors, fusion reactors and high-power large accelerator facilities.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"610 ","pages":"Article 155772"},"PeriodicalIF":2.8,"publicationDate":"2025-03-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143785554","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}
Surface blistering and internal microstructure evolutions as well as deuterium retention in tungsten with helium ion implanted followed by deuterium plasma exposure were investigated. The helium ion implantation was taken with 40 keV with a flux of 1.6 × 1017 He+/(m2s) to a fluence of 6.0 × 1020 He+/m2 at room temperature. The following deuterium plasma exposure was taken with a flux of 5.96 × 1019 D/(m2s) at a bias of 100 eV at 340 K. The deuterium plasma exposure was designed with two different durations. One is about 19 h (h) which corresponds a fluence of 4.07 × 1024 D/m2, while another is nearly 96 h corresponds a fluence of 2.06 × 1025 D/m2. The helium ion implantation itself did not induce surface blister nor detectable internal helium bubble. After subsequent deuterium exposure of 19 h, dense surface blisters appeared on the reference tungsten, while no blister was formed on the helium implanted tungsten, indicating the helium ion implantation can efficiently suppress the surface blistering. However, when the deuterium irradiation time was increased up to 96 h, sparse deuterium blisters appeared on the surface of the helium ion pre-implanted W, indicating D could pass through the helium implantation layer as the exposure time was long enough. TEM results revealed that no bubble can be observed in the reference tungsten only exposed to deuterium plasma, while bubbles can be confirmed in the helium ion pre-implanted tungsten after deuterium irradiation, suggesting that the growth of helium bubbles can be enhanced by the subsequent deuterium plasma exposure. For the deuterium plasma exposure with 19 h, the total deuterium retention in the helium ion pre-implanted tungsten was three times that of the reference tungsten, indicating the helium ion implantation could increase the deuterium retention in tungsten.
{"title":"Impact of helium ion implantation on deuterium plasma induced microstructure evolution and deuterium retention in tungsten","authors":"Honghui Zhang , Tongjun Xia , Yongzhi Shi , Zhengyu Jiang , Xingyu Ren , Lisha Liang , Kaigui Zhu","doi":"10.1016/j.jnucmat.2025.155794","DOIUrl":"10.1016/j.jnucmat.2025.155794","url":null,"abstract":"<div><div>Surface blistering and internal microstructure evolutions as well as deuterium retention in tungsten with helium ion implanted followed by deuterium plasma exposure were investigated. The helium ion implantation was taken with 40 keV with a flux of 1.6 × 10<sup>17</sup> He<sup>+</sup>/(m<sup>2</sup>s) to a fluence of 6.0 × 10<sup>20</sup> He<sup>+</sup>/m<sup>2</sup> at room temperature. The following deuterium plasma exposure was taken with a flux of 5.96 × 10<sup>19</sup> D/(m<sup>2</sup>s) at a bias of 100 eV at 340 K. The deuterium plasma exposure was designed with two different durations. One is about 19 h (h) which corresponds a fluence of 4.07 × 10<sup>24</sup> D/m<sup>2</sup>, while another is nearly 96 h corresponds a fluence of 2.06 × 10<sup>25</sup> D/m<sup>2</sup>. The helium ion implantation itself did not induce surface blister nor detectable internal helium bubble. After subsequent deuterium exposure of 19 h, dense surface blisters appeared on the reference tungsten, while no blister was formed on the helium implanted tungsten, indicating the helium ion implantation can efficiently suppress the surface blistering. However, when the deuterium irradiation time was increased up to 96 h, sparse deuterium blisters appeared on the surface of the helium ion pre-implanted W, indicating D could pass through the helium implantation layer as the exposure time was long enough. TEM results revealed that no bubble can be observed in the reference tungsten only exposed to deuterium plasma, while bubbles can be confirmed in the helium ion pre-implanted tungsten after deuterium irradiation, suggesting that the growth of helium bubbles can be enhanced by the subsequent deuterium plasma exposure. For the deuterium plasma exposure with 19 h, the total deuterium retention in the helium ion pre-implanted tungsten was three times that of the reference tungsten, indicating the helium ion implantation could increase the deuterium retention in tungsten.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"610 ","pages":"Article 155794"},"PeriodicalIF":2.8,"publicationDate":"2025-03-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143777488","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}
The dissolution of International Simple Glass (ISG) was investigated at 90 °C, elevated concentration of dissolved silica and in the presence of calcium, with a specific emphasis on basic pH conditions. The leaching solution was labelled with 29Si, 18O and 44Ca in part of the experiments to elucidate the dissolution mechanisms. Based on the isotopic signatures of the gel layer analyzed using Time-of-Flight Secondary Ion Mass Spectrometry (ToF-SIMS), it was concluded that oxygen atoms mostly originate from the solution for all investigated conditions, while silicon atoms almost exclusively originate from the glass. A negative correlation was found between the initial concentration of calcium in solution and the gel layer thickness, suggesting either the formation of a passivating (Si, Ca)-rich layer, a catalytic effect of Ca on the gel densification or a combination of both. In addition, the pH-dependence of the diffusion coefficient of B within the gel was found to be stronger in the basic pH range than in the acidic pH range, which was suggested to originate from the change in coordination of B species at pH90 °C ∼ 8.5. Overall, these results suggest that in a (Ca, Si)-rich solution at basic pH, the durability of ISG is stronger than previously thought, as the diffusion coefficient of B under such conditions are lower than expected based on literature.
{"title":"Impact of calcium and pH on ISG alteration at basic pH: Mechanism of formation and transport properties of the gel layer","authors":"Benjamin Cagnon , Stéphane Gin , Martiane Cabié , Damien Daval","doi":"10.1016/j.jnucmat.2025.155796","DOIUrl":"10.1016/j.jnucmat.2025.155796","url":null,"abstract":"<div><div>The dissolution of International Simple Glass (ISG) was investigated at 90 °C, elevated concentration of dissolved silica and in the presence of calcium, with a specific emphasis on basic pH conditions. The leaching solution was labelled with <sup>29</sup>Si, <sup>18</sup>O and <sup>44</sup>Ca in part of the experiments to elucidate the dissolution mechanisms. Based on the isotopic signatures of the gel layer analyzed using Time-of-Flight Secondary Ion Mass Spectrometry (ToF-SIMS), it was concluded that oxygen atoms mostly originate from the solution for all investigated conditions, while silicon atoms almost exclusively originate from the glass. A negative correlation was found between the initial concentration of calcium in solution and the gel layer thickness, suggesting either the formation of a passivating (Si, Ca)-rich layer, a catalytic effect of Ca on the gel densification or a combination of both. In addition, the pH-dependence of the diffusion coefficient of B within the gel was found to be stronger in the basic pH range than in the acidic pH range, which was suggested to originate from the change in coordination of B species at pH<sub>90</sub> °<sub>C</sub> ∼ 8.5. Overall, these results suggest that in a (Ca, Si)-rich solution at basic pH, the durability of ISG is stronger than previously thought, as the diffusion coefficient of B under such conditions are lower than expected based on literature.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"610 ","pages":"Article 155796"},"PeriodicalIF":2.8,"publicationDate":"2025-03-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143768659","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}
The impact of substituting Mn for Ni on the microstructure and tensile properties of alumina-forming austenitic (AFA) stainless steel was systematically studied. The findings revealed that the addition of 4 wt. % Mn in place of 2 wt. % Ni could inhibit the precipitation of the B2-NiAl phase but increase the aspect ratio of the B2-NiAl particles and promote the precipitation of the Laves phase. The addition of Mn also promoted the formation of coincidence site lattice (CSL) grain boundaries and Goss texture, therefore beneficial for improving the mechanical properties. After aging at 700 °C, the room temperature (RT) ultimate tensile strength (UTS) and elongation of Mn-added AFA steel significantly improved to 1037.5 MPa and 34.53 %, respectively, compared to the Mn-free AFA steel, which exhibited a UTS of 848.35 MPa and elongation of 26.4 %. Notably, when tested at 700 °C, the elongation of Mn-added steel reached 60.5 %, nearly double that of Mn-free steel (36.5 %), while maintaining similar strength.
{"title":"Effect of Mn replacing Ni on the microstructure and tensile properties of alumina-forming austenitic stainless steel","authors":"Guoshuai Chen , Shang Du , Lingzhi Chen , Weiwei Cong , Zhangjian Zhou","doi":"10.1016/j.jnucmat.2025.155785","DOIUrl":"10.1016/j.jnucmat.2025.155785","url":null,"abstract":"<div><div>The impact of substituting Mn for Ni on the microstructure and tensile properties of alumina-forming austenitic (AFA) stainless steel was systematically studied. The findings revealed that the addition of 4 wt. % Mn in place of 2 wt. % Ni could inhibit the precipitation of the B2-NiAl phase but increase the aspect ratio of the B2-NiAl particles and promote the precipitation of the Laves phase. The addition of Mn also promoted the formation of coincidence site lattice (CSL) grain boundaries and Goss texture, therefore beneficial for improving the mechanical properties. After aging at 700 °C, the room temperature (RT) ultimate tensile strength (UTS) and elongation of Mn-added AFA steel significantly improved to 1037.5 MPa and 34.53 %, respectively, compared to the Mn-free AFA steel, which exhibited a UTS of 848.35 MPa and elongation of 26.4 %. Notably, when tested at 700 °C, the elongation of Mn-added steel reached 60.5 %, nearly double that of Mn-free steel (36.5 %), while maintaining similar strength.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"610 ","pages":"Article 155785"},"PeriodicalIF":2.8,"publicationDate":"2025-03-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143768658","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 : 2025-03-27DOI: 10.1016/j.jnucmat.2025.155783
Mengxian Xiang , Ning Ma , Weiquan Liang , Yinsong Xie , Sizhi Zuo-Jiang , Xuzhou Jiang , Hongying Yu , Dongbai Sun
The radioactive nature of uranium and the flammability of uranium hydride make it difficult to monitor and characterize its corrosion state in real time, limiting the study of the corrosion behavior of uranium in a hydrogen environment. In this work, a three-dimensional transient corrosion model of uranium metal material in hydrogen environment is established based on the finite element method, which intuitively reveals the corrosion behavior and corrosion mechanism of the hydride incubation, nucleation and growth process of uranium in hydrogen environment. The formation of hydride process of uranium undergoes three stages: the incubation stage, the nucleation stage, and the acceleration stage. The corrosion primarily manifests as pitting corrosion, progressing in a spherical morphology. The uranium-hydrogen reaction incubation stage arises mainly from the process of hydrogen diffusion. The strong stress accumulation and damage to the matrix induced by the volume expansion of hydride is the root origin of the formation of hydrogen pits. The accelerating effect of temperature on the reaction rate of uranium-hydrogen mainly results from the increase in the reaction rate constant and the significant increase in the diffusion rate. The magnitude of the elastic modulus and Poisson's ratio of uranium is positively correlated with the stress accumulation caused by hydride growth and the degree of damage to the uranium matrix. The findings provide a basis for designing new corrosion-resistant uranium alloys through numerical simulation. This work offers valuable insights into the prediction of damage caused by hydride formation in hydrogen-containing environment.
{"title":"Pitting corrosion simulation from UH3 expansion in uranium hydrogen environment with a validation through existing experimental data","authors":"Mengxian Xiang , Ning Ma , Weiquan Liang , Yinsong Xie , Sizhi Zuo-Jiang , Xuzhou Jiang , Hongying Yu , Dongbai Sun","doi":"10.1016/j.jnucmat.2025.155783","DOIUrl":"10.1016/j.jnucmat.2025.155783","url":null,"abstract":"<div><div>The radioactive nature of uranium and the flammability of uranium hydride make it difficult to monitor and characterize its corrosion state in real time, limiting the study of the corrosion behavior of uranium in a hydrogen environment. In this work, a three-dimensional transient corrosion model of uranium metal material in hydrogen environment is established based on the finite element method, which intuitively reveals the corrosion behavior and corrosion mechanism of the hydride incubation, nucleation and growth process of uranium in hydrogen environment. The formation of hydride process of uranium undergoes three stages: the incubation stage, the nucleation stage, and the acceleration stage. The corrosion primarily manifests as pitting corrosion, progressing in a spherical morphology. The uranium-hydrogen reaction incubation stage arises mainly from the process of hydrogen diffusion. The strong stress accumulation and damage to the matrix induced by the volume expansion of hydride is the root origin of the formation of hydrogen pits. The accelerating effect of temperature on the reaction rate of uranium-hydrogen mainly results from the increase in the reaction rate constant and the significant increase in the diffusion rate. The magnitude of the elastic modulus and Poisson's ratio of uranium is positively correlated with the stress accumulation caused by hydride growth and the degree of damage to the uranium matrix. The findings provide a basis for designing new corrosion-resistant uranium alloys through numerical simulation. This work offers valuable insights into the prediction of damage caused by hydride formation in hydrogen-containing environment.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"610 ","pages":"Article 155783"},"PeriodicalIF":2.8,"publicationDate":"2025-03-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143799585","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 : 2025-03-26DOI: 10.1016/j.jnucmat.2025.155782
Allison Probert , Alexander Swearingen , Jason Schulthess , Luca Capriotti , Colby Jensen , Assel Aitkaliyeva
The development of next-generation sodium-cooled fast reactors necessitates comprehensive research on metallic fuels to maximize economic performance while ensuring safe operation. In this study, we investigated the steady-state irradiation behavior of two high burnup U-19Pu-10Zr fuel pins, DP-36 and DP-40, in preparation for planned safety testing. Post-irradiation examination (PIE) was performed to quantify fuel column elongation, regions of low-density at the top of the fuel column, pin deformation, fission product distribution, fractional fission gas release, microstructural evolution, and fuel constituent redistribution. Benchmarking against existing PIE data from U-19Pu-10Zr fuel pins irradiated in EBR-II revealed consistent patterns in fuel column elongation and cladding diametral strain. However, both pins exhibited longer low-density structures, and destructive examination of DP-36 revealed more complex constituent redistribution patterns compared to previously reported data for ternary fuel pins. The steady-state irradiation of both pins was also modeled using BISON. Comparisons of PIE results with modeled predictions showed overall agreement in fractional fission gas release but consistent overestimation of axial and radial swelling due to gaseous and solid swelling models. These findings underscore the critical importance of pre-test characterization on test and sibling pins to accurately capture steady-state fuel behavior ahead of transient testing, thus establishing a baseline for post-test comparison. Additionally, these analyses identified key data gaps that warrant further investigation to improve the understanding and prediction of fuel swelling, thereby enhancing the synergy between modeling and experimental efforts in supporting accident testing.
{"title":"Comparative post-irradiation examination of high burnup U-19Pu-10Zr: Assessing steady-state irradiation behavior against historical and modeled fuel performance","authors":"Allison Probert , Alexander Swearingen , Jason Schulthess , Luca Capriotti , Colby Jensen , Assel Aitkaliyeva","doi":"10.1016/j.jnucmat.2025.155782","DOIUrl":"10.1016/j.jnucmat.2025.155782","url":null,"abstract":"<div><div>The development of next-generation sodium-cooled fast reactors necessitates comprehensive research on metallic fuels to maximize economic performance while ensuring safe operation. In this study, we investigated the steady-state irradiation behavior of two high burnup U-19Pu-10Zr fuel pins, DP-36 and DP-40, in preparation for planned safety testing. Post-irradiation examination (PIE) was performed to quantify fuel column elongation, regions of low-density at the top of the fuel column, pin deformation, fission product distribution, fractional fission gas release, microstructural evolution, and fuel constituent redistribution. Benchmarking against existing PIE data from U-19Pu-10Zr fuel pins irradiated in EBR-II revealed consistent patterns in fuel column elongation and cladding diametral strain. However, both pins exhibited longer low-density structures, and destructive examination of DP-36 revealed more complex constituent redistribution patterns compared to previously reported data for ternary fuel pins. The steady-state irradiation of both pins was also modeled using BISON. Comparisons of PIE results with modeled predictions showed overall agreement in fractional fission gas release but consistent overestimation of axial and radial swelling due to gaseous and solid swelling models. These findings underscore the critical importance of pre-test characterization on test and sibling pins to accurately capture steady-state fuel behavior ahead of transient testing, thus establishing a baseline for post-test comparison. Additionally, these analyses identified key data gaps that warrant further investigation to improve the understanding and prediction of fuel swelling, thereby enhancing the synergy between modeling and experimental efforts in supporting accident testing.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"610 ","pages":"Article 155782"},"PeriodicalIF":2.8,"publicationDate":"2025-03-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143785494","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 : 2025-03-26DOI: 10.1016/j.jnucmat.2025.155781
Qinzeng Hu , Lingyan Xu , Zhixin Tan , Ming Hao , Lu Liang , Yingming Wang , Zhentao Qin , Lixiang Lian , Chongqi Liu , Yanyan Lei , Wei Zheng , Wanqi Jie
Semiconductor radiation detectors used in nuclear power plants and other environments are inevitably exposed to neutron, γ-ray and other high-energy radiation, which can damage the crystal structure of semiconductors and thus degrade the detector performance. Here, we investigate the effects of neutron irradiation on the microstructure, photoelectric and radiation detection performance of CdZnTe detectors. Low-temperature photoluminescence (PL) spectra show that the dislocation related defect concentration in the irradiated crystals increases with increasing fluence. The infrared (IR) transmittance of the irradiated crystal decreases compared with that of the unirradiated crystal, which also indicates an increase in the dislocation density. The presence of stacking faults, stacking fault dipoles and dislocation locks in the irradiated CdZnTe crystals has been revealed by transmission electron microscopy (TEM). The energy resolution of γ-ray from 241Am@100 V is degraded from 5.86 % before irradiation to 10.72 % after irradiation at 5.6 × 1010 n/cm2. In addition, the mobility-lifetime product of electron (μτ)e in CdZnTe detectors is reduced from 4.8 × 10-3 cm2/V before irradiation to 7.02 × 10-4 cm2/V after irradiation at 5.6 × 1010 n/cm2. I-V test show that the barrier height of the CdZnTe detector decreases with the increase of neutron irradiation fluence, leading to a decrease in resistivity. Time-of-flight (TOF) tests demonstrate that the electron mobility after irradiation decreases with increasing irradiation fluence. Notably, the maximum neutron fluence used in this study is 3.9 × 1011 n/cm2, at which the CdZnTe radiation detector is not completely damaged. This study mainly investigates the radiation damage mechanism, induced defect characteristics and performance degradation of CdZnTe crystals by neutron irradiation, aiming to provide theoretical guidance for improving the radiation-resistant properties of detectors.
{"title":"Damage generation mechanism and performance degradation of CdZnTe radiation detectors in neutron radiation field","authors":"Qinzeng Hu , Lingyan Xu , Zhixin Tan , Ming Hao , Lu Liang , Yingming Wang , Zhentao Qin , Lixiang Lian , Chongqi Liu , Yanyan Lei , Wei Zheng , Wanqi Jie","doi":"10.1016/j.jnucmat.2025.155781","DOIUrl":"10.1016/j.jnucmat.2025.155781","url":null,"abstract":"<div><div>Semiconductor radiation detectors used in nuclear power plants and other environments are inevitably exposed to neutron, γ-ray and other high-energy radiation, which can damage the crystal structure of semiconductors and thus degrade the detector performance. Here, we investigate the effects of neutron irradiation on the microstructure, photoelectric and radiation detection performance of CdZnTe detectors. Low-temperature photoluminescence (PL) spectra show that the dislocation related defect concentration in the irradiated crystals increases with increasing fluence. The infrared (IR) transmittance of the irradiated crystal decreases compared with that of the unirradiated crystal, which also indicates an increase in the dislocation density. The presence of stacking faults, stacking fault dipoles and dislocation locks in the irradiated CdZnTe crystals has been revealed by transmission electron microscopy (TEM). The energy resolution of γ-ray from <sup>241</sup>Am@100 V is degraded from 5.86 % before irradiation to 10.72 % after irradiation at 5.6 × 10<sup>10</sup> n/cm<sup>2</sup>. In addition, the mobility-lifetime product of electron (μτ)<sub>e</sub> in CdZnTe detectors is reduced from 4.8 × 10<sup>-3</sup> cm<sup>2</sup>/V before irradiation to 7.02 × 10<sup>-4</sup> cm<sup>2</sup>/V after irradiation at 5.6 × 10<sup>10</sup> n/cm<sup>2</sup>. I-V test show that the barrier height of the CdZnTe detector decreases with the increase of neutron irradiation fluence, leading to a decrease in resistivity. Time-of-flight (TOF) tests demonstrate that the electron mobility after irradiation decreases with increasing irradiation fluence. Notably, the maximum neutron fluence used in this study is 3.9 × 10<sup>11</sup> n/cm<sup>2</sup>, at which the CdZnTe radiation detector is not completely damaged. This study mainly investigates the radiation damage mechanism, induced defect characteristics and performance degradation of CdZnTe crystals by neutron irradiation, aiming to provide theoretical guidance for improving the radiation-resistant properties of detectors.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"610 ","pages":"Article 155781"},"PeriodicalIF":2.8,"publicationDate":"2025-03-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143734527","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 : 2025-03-25DOI: 10.1016/j.jnucmat.2025.155780
Okan Yetik , Pavel Trtik , Robert Zubler , Robin Manuel Grabherr , Johannes Bertsch , Liliana I. Duarte
Duplex claddings, a concept with an outer layer on Zircaloy-4 substrate, have been employed in Swiss pressurized water reactors for decades. A significant amount of spent fuel encapsulated within this cladding type is currently stored in interim dry storage facilities before relocation to deep geological repositories. Duplex claddings exhibit different hydrogen distribution behaviour compared to single material claddings. This behaviour has been previously observed in non-irradiated duplex claddings. In this study, we extended the investigation to reactor-irradiated duplex cladding with an average burnup of 71.6 MWd/kg (HM). The results show that chemical potential and terminal solid solubility influence hydrogen distribution in both non-irradiated and reactor-irradiated cladding. However, the initial distribution of hydrogen plays a significant role in its subsequent redistribution. The formation of a dense hydride accumulation zone in the liner (DHAZliner) is evident and arises from differences in hydrogen solubility between the liner and substrate. Irradiated claddings demonstrate higher hydrogen mobility within the cladding, making them more responsive to external stresses in comparison to non-irradiated claddings.
{"title":"Hydrogen redistribution in non-irradiated and irradiated duplex zirconium claddings by high-resolution neutron imaging","authors":"Okan Yetik , Pavel Trtik , Robert Zubler , Robin Manuel Grabherr , Johannes Bertsch , Liliana I. Duarte","doi":"10.1016/j.jnucmat.2025.155780","DOIUrl":"10.1016/j.jnucmat.2025.155780","url":null,"abstract":"<div><div>Duplex claddings, a concept with an outer layer on Zircaloy-4 substrate, have been employed in Swiss pressurized water reactors for decades. A significant amount of spent fuel encapsulated within this cladding type is currently stored in interim dry storage facilities before relocation to deep geological repositories. Duplex claddings exhibit different hydrogen distribution behaviour compared to single material claddings. This behaviour has been previously observed in non-irradiated duplex claddings. In this study, we extended the investigation to reactor-irradiated duplex cladding with an average burnup of 71.6 MWd/kg (HM). The results show that chemical potential and terminal solid solubility influence hydrogen distribution in both non-irradiated and reactor-irradiated cladding. However, the initial distribution of hydrogen plays a significant role in its subsequent redistribution. The formation of a dense hydride accumulation zone in the liner (DHAZ<sub>liner</sub>) is evident and arises from differences in hydrogen solubility between the liner and substrate. Irradiated claddings demonstrate higher hydrogen mobility within the cladding, making them more responsive to external stresses in comparison to non-irradiated claddings.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"610 ","pages":"Article 155780"},"PeriodicalIF":2.8,"publicationDate":"2025-03-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143777575","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 : 2025-03-25DOI: 10.1016/j.jnucmat.2025.155767
M. Baxter Chinery , H.S. Wong , M.R. Wenman , C.R. Cheeseman , L.J. Vandeperre
A novel approach to reduce the volume of the UK legacy Magnox waste is to use the Mg(OH)2-rich sludge as a component in a magnesium-silicate-hydrate (M-S-H) mortar. This study looks at the potential to incorporate a simulant corroded magnesium sludge (CMgS) into a mortar that achieves the encapsulation criterion. The optimum M-S-H mortar has a binder molar ratio of 0.5, resulting in 46.6 wt% of the mortar being CMgS sludge. This mortar had compressive strengths of 25 MPa at 60 days with a connected porosity of 6.5 % at 6 months, and this meets the requirements for a waste encapsulated grout. The work demonstrates the potential to use a Mg(OH)2-rich sludge, obtained from Magnox Swarf Storage Silos as a raw material in M-S-H mortars in waste encapsulation, with associated cost and space savings.
{"title":"Development of magnesium-silicate-hydrate mortars using magnesium hydroxide for Magnox waste encapsulation","authors":"M. Baxter Chinery , H.S. Wong , M.R. Wenman , C.R. Cheeseman , L.J. Vandeperre","doi":"10.1016/j.jnucmat.2025.155767","DOIUrl":"10.1016/j.jnucmat.2025.155767","url":null,"abstract":"<div><div>A novel approach to reduce the volume of the UK legacy Magnox waste is to use the Mg(OH)<sub>2</sub>-rich sludge as a component in a magnesium-silicate-hydrate (M-S-H) mortar. This study looks at the potential to incorporate a simulant corroded magnesium sludge (CMgS) into a mortar that achieves the encapsulation criterion. The optimum M-S-H mortar has a binder molar ratio of 0.5, resulting in 46.6 wt% of the mortar being CMgS sludge. This mortar had compressive strengths of 25 MPa at 60 days with a connected porosity of 6.5 % at 6 months, and this meets the requirements for a waste encapsulated grout. The work demonstrates the potential to use a Mg(OH)<sub>2</sub>-rich sludge, obtained from Magnox Swarf Storage Silos as a raw material in M-S-H mortars in waste encapsulation, with associated cost and space savings.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"610 ","pages":"Article 155767"},"PeriodicalIF":2.8,"publicationDate":"2025-03-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143785555","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}
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