Pub Date : 2025-02-01DOI: 10.1016/j.jnucmat.2024.155557
Jaewoo Park, Jinsuo Zhang
{"title":"Corrigendum to “Non-electric mass transfer between stainless steel 316H and glassy carbon in NaF-KF-UF4 salt” [Journal of Nuclear Materials 604 (2025) 155534]","authors":"Jaewoo Park, Jinsuo Zhang","doi":"10.1016/j.jnucmat.2024.155557","DOIUrl":"10.1016/j.jnucmat.2024.155557","url":null,"abstract":"","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"605 ","pages":"Article 155557"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143170987","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 : 2025-02-01DOI: 10.1016/j.jnucmat.2024.155573
Yubo Zhou , Yingxuan Dong , Haojun Ma , Junnan Lv , Qun Li
For nuclear fuel elements, the interface mechanical properties of zirconium alloy cladding is critical to the safety and reliability of reactors. However, due to the small thickness of the fuel plates (<2 mm), accurately capturing the behavior of interface cracks is challenging, complicates the measurement of interface fracture toughness. This study developed a data-driven inversion method using the generalized regression neural network (GRNN) to rapidly and accurately determine the fracture toughness of zirconium alloy cladding interface. The database was established by combining splitting experiments with numerical simulations. The cohesive zone model was utilized to accurately simulate crack propagation paths and fracture modes in numerical simulations. The influences of key parameters such as cohesive strength, stiffness, and interface fracture energy were analyzed in detail. After extensive training, the prediction model accurately forecasted the interface fracture toughness. The results indicate that the proposed GRNN-based inversion approach is feasible and effective for predicting the fracture toughness of zirconium alloy cladding interface, and can be extended to determinations of other mechanical properties in the nuclear fuel element.
{"title":"Inversion of the fracture toughness of zirconium alloy cladding interface in nuclear fuel using splitting method via general regression neural network","authors":"Yubo Zhou , Yingxuan Dong , Haojun Ma , Junnan Lv , Qun Li","doi":"10.1016/j.jnucmat.2024.155573","DOIUrl":"10.1016/j.jnucmat.2024.155573","url":null,"abstract":"<div><div>For nuclear fuel elements, the interface mechanical properties of zirconium alloy cladding is critical to the safety and reliability of reactors. However, due to the small thickness of the fuel plates (<2 mm), accurately capturing the behavior of interface cracks is challenging, complicates the measurement of interface fracture toughness. This study developed a data-driven inversion method using the generalized regression neural network (GRNN) to rapidly and accurately determine the fracture toughness of zirconium alloy cladding interface. The database was established by combining splitting experiments with numerical simulations. The cohesive zone model was utilized to accurately simulate crack propagation paths and fracture modes in numerical simulations. The influences of key parameters such as cohesive strength, stiffness, and interface fracture energy were analyzed in detail. After extensive training, the prediction model accurately forecasted the interface fracture toughness. The results indicate that the proposed GRNN-based inversion approach is feasible and effective for predicting the fracture toughness of zirconium alloy cladding interface, and can be extended to determinations of other mechanical properties in the nuclear fuel element.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"605 ","pages":"Article 155573"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143171593","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 : 2025-02-01DOI: 10.1016/j.jnucmat.2024.155537
M. Poschmann, D. Wojtaszek, E. Geiger
Highly-detailed equilibrium thermodynamics calculations were used to derive bounds on safe fresh UCO TRISO fuel kernel chemistries, in terms of the fraction of UCx. The lower bound on UCx was chosen to limit CO pressure, and the upper bound to ensure oxidation of radionuclide fission and transmutation products of concern. The bounds can be calculated with respect to both target burnup and fuel temperature. Calculated bounds for a demonstration fuel design similar to the AGR-2 experiment are in good agreement with previously-calculated lower bounds and experimental evidence used to determine upper bounds, and indicate that reference fuel designs such as AGR-2 and MHTGR-350 are well-within the safe composition region even for burnup as high as 21.3% FIMA. Predicted speciations of cesium, palladium, silver, barium, strontium, and zirconium were examined, and recommendations for thermodynamic assessments of relevant systems provided.
{"title":"Thermochemical bounds on UCO TRISO kernel compositions","authors":"M. Poschmann, D. Wojtaszek, E. Geiger","doi":"10.1016/j.jnucmat.2024.155537","DOIUrl":"10.1016/j.jnucmat.2024.155537","url":null,"abstract":"<div><div>Highly-detailed equilibrium thermodynamics calculations were used to derive bounds on safe fresh UCO TRISO fuel kernel chemistries, in terms of the fraction of UC<sub><em>x</em></sub>. The lower bound on UC<sub><em>x</em></sub> was chosen to limit CO pressure, and the upper bound to ensure oxidation of radionuclide fission and transmutation products of concern. The bounds can be calculated with respect to both target burnup and fuel temperature. Calculated bounds for a demonstration fuel design similar to the AGR-2 experiment are in good agreement with previously-calculated lower bounds and experimental evidence used to determine upper bounds, and indicate that reference fuel designs such as AGR-2 and MHTGR-350 are well-within the safe composition region even for burnup as high as 21.3% FIMA. Predicted speciations of cesium, palladium, silver, barium, strontium, and zirconium were examined, and recommendations for thermodynamic assessments of relevant systems provided.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"605 ","pages":"Article 155537"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143171598","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 : 2025-02-01DOI: 10.1016/j.jnucmat.2025.155608
Xinrun Chen , Tatsuya Suzuki , Phongsakorn Prak Tom , Bo Li , Zongda Yang , Sho Kano , Takuya Yamamoto , Kenta Murakami
The present study investigates the irradiation hardening in pure Fe and Fe-0.3 wt.% Si alloys after He+ ion implantation tests in order to obtain a representative value for the barrier strength (α-value) of helium bubbles in model ferritic alloys. 160 keV He+ ion implantation was performed at 400 ± 2 °C with a fluence up to 3.6 × 1016 cm-2 at a flux of 4.0 × 1013 cm-2∙s-1. Hardness was measured using the nanoindentation technique at different indentation depths, and the depth-distributed helium bubbles were observed by transmission electron microscopy (TEM). The results of the hardness measurement demonstrated a significant damage gradient effect in both He+ implanted specimens. This could be attributed to the presence of depth-distributed helium bubbles. Additionally, a significant dose dependence on swelling was observed in both alloys, suggesting that the primary parameter governing the evolution of helium bubbles is the dose. While the traditional dispersed barrier hardening (DBH) model could be employed to evaluate the α-value based on helium bubble-induced hardening, the distribution of dispersed barriers should be uniform. Consequently, we proposed a more efficacious application of the dispersed barrier hardening model to describe the depth-distributed helium bubble-induced hardening in iron-based materials by integrating the nanoindentation technique with TEM. The α-value for helium bubbles of approximately 2∼3 nm in diameter in model ferritic alloys is estimated to be approximately 0.15 to 0.16 based on the revised DBH model, which is consistent with the value observed in annealed F82H of neutron irradiation. Furthermore, this study suggested that the much greater hardness increase and swelling rate of implanted Fe-0.3Si than that of pure iron is due to the additive silicon, which has a strong ability to inhibit vacancy diffusion in the evolution of helium bubbles.
{"title":"Dispersed barrier hardening modeling on depth-distributed helium bubbles in iron-based alloys","authors":"Xinrun Chen , Tatsuya Suzuki , Phongsakorn Prak Tom , Bo Li , Zongda Yang , Sho Kano , Takuya Yamamoto , Kenta Murakami","doi":"10.1016/j.jnucmat.2025.155608","DOIUrl":"10.1016/j.jnucmat.2025.155608","url":null,"abstract":"<div><div>The present study investigates the irradiation hardening in pure Fe and Fe-0.3 wt.% Si alloys after He<sup>+</sup> ion implantation tests in order to obtain a representative value for the barrier strength (α-value) of helium bubbles in model ferritic alloys. 160 keV He<sup>+</sup> ion implantation was performed at 400 ± 2 °C with a fluence up to 3.6 × 10<sup>16</sup> cm<sup>-2</sup> at a flux of 4.0 × 10<sup>13</sup> cm<sup>-2</sup>∙s<sup>-1</sup>. Hardness was measured using the nanoindentation technique at different indentation depths, and the depth-distributed helium bubbles were observed by transmission electron microscopy (TEM). The results of the hardness measurement demonstrated a significant damage gradient effect in both He<sup>+</sup> implanted specimens. This could be attributed to the presence of depth-distributed helium bubbles. Additionally, a significant dose dependence on swelling was observed in both alloys, suggesting that the primary parameter governing the evolution of helium bubbles is the dose. While the traditional dispersed barrier hardening (DBH) model could be employed to evaluate the α-value based on helium bubble-induced hardening, the distribution of dispersed barriers should be uniform. Consequently, we proposed a more efficacious application of the dispersed barrier hardening model to describe the depth-distributed helium bubble-induced hardening in iron-based materials by integrating the nanoindentation technique with TEM. The α-value for helium bubbles of approximately 2∼3 nm in diameter in model ferritic alloys is estimated to be approximately 0.15 to 0.16 based on the revised DBH model, which is consistent with the value observed in annealed F82H of neutron irradiation. Furthermore, this study suggested that the much greater hardness increase and swelling rate of implanted Fe-0.3Si than that of pure iron is due to the additive silicon, which has a strong ability to inhibit vacancy diffusion in the evolution of helium bubbles.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"606 ","pages":"Article 155608"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143154878","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 : 2025-02-01DOI: 10.1016/j.jnucmat.2025.155618
Danny J. Edwards , Alan Schemer-Kohrn , Matt Olszta , Ramprashad Prabhakaran , Yuanyuan Zhu , Jing Wang , Jacob Haag , Osman El Atwani , Timothy G. Lach , Mychailo Toloczko
Focused ion beam methods have often become a de facto choice for metallic materials that could also be easily electropolished. FIB preparation of TEM samples is widely used for radioactive materials, since the site specificity, coupled with the minuscule volume of material used for the TEM sample, can produce TEM samples with minimal to no detectable radioactivity. FIB preparation has problems however, as evidenced by artifacts such the subsurface black spots (clusters of vacancies and/or interstitials), dislocations, amorphous layers and phase changes, and must be accounted for when conducting TEM experiments. This study has two main objectives, first, presenting evidence on two types of surface artifacts (moiré fringes and surface dislocations) observed in FIB prepared Fe-Cr alloys and pure Fe. This evidence will include similar artifacts produced in ferritic based systems using two other ion sputtering techniques: conventional Ar+ ion milling with a Gatan PIPS®, and an Ar+ based Fischione NanoMill®. The second objective documents our method of removing all of the FIB artifacts using flash electropolishing (FEP) of FIB TEM lamella, demonstrating our success in producing electropolished FIB lamella from an FeCr HT-9 alloy free of both the subsurface and surface artifacts. With the proper choice of parameters, TEM samples from FeCr alloys are comparable to those prepared by jet polishing of bulk TEM samples. These comparisons between FIB prepared and FEP FIB samples are done using both TEM imaging and diffraction contrast imaging STEM (DCI-STEM) on samples from an unirradiated and neutron irradiated Fe-Cr alloy. The observed black spot damage, moiré fringes and surface dislocations in the ion beam prepared samples are discussed in terms of how they could impact the microstructural analysis of irradiated metals.
{"title":"Understanding and removing FIB artifacts in metallic TEM samples using flash electropolishing","authors":"Danny J. Edwards , Alan Schemer-Kohrn , Matt Olszta , Ramprashad Prabhakaran , Yuanyuan Zhu , Jing Wang , Jacob Haag , Osman El Atwani , Timothy G. Lach , Mychailo Toloczko","doi":"10.1016/j.jnucmat.2025.155618","DOIUrl":"10.1016/j.jnucmat.2025.155618","url":null,"abstract":"<div><div>Focused ion beam methods have often become a de facto choice for metallic materials that could also be easily electropolished. FIB preparation of TEM samples is widely used for radioactive materials, since the site specificity, coupled with the minuscule volume of material used for the TEM sample, can produce TEM samples with minimal to no detectable radioactivity. FIB preparation has problems however, as evidenced by artifacts such the subsurface black spots (clusters of vacancies and/or interstitials), dislocations, amorphous layers and phase changes, and must be accounted for when conducting TEM experiments. This study has two main objectives, first, presenting evidence on two types of surface artifacts (moiré fringes and surface dislocations) observed in FIB prepared Fe-Cr alloys and pure Fe. This evidence will include similar artifacts produced in ferritic based systems using two other ion sputtering techniques: conventional Ar<sup>+</sup> ion milling with a Gatan PIPS®, and an Ar<sup>+</sup> based Fischione NanoMill®. The second objective documents our method of removing all of the FIB artifacts using flash electropolishing (FEP) of FIB TEM lamella, demonstrating our success in producing electropolished FIB lamella from an FeCr HT-9 alloy free of both the subsurface and surface artifacts. With the proper choice of parameters, TEM samples from FeCr alloys are comparable to those prepared by jet polishing of bulk TEM samples. These comparisons between FIB prepared and FEP FIB samples are done using both TEM imaging and diffraction contrast imaging STEM (DCI-STEM) on samples from an unirradiated and neutron irradiated Fe-Cr alloy. The observed black spot damage, moiré fringes and surface dislocations in the ion beam prepared samples are discussed in terms of how they could impact the microstructural analysis of irradiated metals.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"606 ","pages":"Article 155618"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143154921","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 : 2025-02-01DOI: 10.1016/j.jnucmat.2025.155624
Jie Qiu , Ryan D. Hayes , Ho Lun Chan , Haley Williams , Digby D. Macdonald , Raluca O. Scarlat , Djamel Kaoumi , John R. Scully , Peter Hosemann
The effect of cold forging on the microstructure and corrosion behavior of 316L stainless steel (SS) in molten salt was investigated. Static corrosion experiments were performed in FLiNaK (LiF-NaF-KF: 46.5–11.5–42 mol.%) at 600 °C for 50 h in a glove box. The results show that cold forging gives rise to enhanced corrosion of 316L SS in molten fluoride salt due to the increase of crystallographic defects. Based on the potentiodynamic polarization results, the corrosion current density of 50 % cold-forged 316L SS is about 2.1 times larger than that of the as-received 316L SS in molten FLiNaK salt at 600 °C.
{"title":"Effect of cold forging on the microstructure and corrosion behavior of type 316L stainless steel in molten FLiNaK salt","authors":"Jie Qiu , Ryan D. Hayes , Ho Lun Chan , Haley Williams , Digby D. Macdonald , Raluca O. Scarlat , Djamel Kaoumi , John R. Scully , Peter Hosemann","doi":"10.1016/j.jnucmat.2025.155624","DOIUrl":"10.1016/j.jnucmat.2025.155624","url":null,"abstract":"<div><div>The effect of cold forging on the microstructure and corrosion behavior of 316L stainless steel (SS) in molten salt was investigated. Static corrosion experiments were performed in FLiNaK (LiF-NaF-KF: 46.5–11.5–42 mol.%) at 600 °C for 50 h in a glove box. The results show that cold forging gives rise to enhanced corrosion of 316L SS in molten fluoride salt due to the increase of crystallographic defects. Based on the potentiodynamic polarization results, the corrosion current density of 50 % cold-forged 316L SS is about 2.1 times larger than that of the as-received 316L SS in molten FLiNaK salt at 600 °C.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"606 ","pages":"Article 155624"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143155318","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 : 2025-02-01DOI: 10.1016/j.jnucmat.2025.155615
Yulin Wei , Xiaoyue Li , Chenhao Yang , Junxiong Liu , Ping Peng , Min Liu
To enhance the He irradiation-resistance of the MAX phase, the irradiation-induced defect evolution of the Ti3SiC2-based two-component MAX phase has been investigated by First Principles. This clarifies the laws and mechanisms of irradiation damage in different structural two-component MAX phases, which offers a crucial reference for the screening of new nuclear structural materials. It was found that the M-site solid solution promoted the vacancies formation, inhibited interstitial dissolution of He, and reduced the damage to atomic stability caused by He-vacancies. The A-site solid solution decreased the formation energy of antisite defects and enhanced the resistance to damage induced by vacancies. He bubbles are less likely to form because both two-component MAX phases prevent He atoms from migrating and binding within the lattice and interfaces. Furthermore, the degree of lattice distortion affected the degree of property alteration and the tendency, which made the comprehensive performance of (Ti0.5Ta0.5)3SiC2 and Ti3(Si0.5Al0.5)C2 better. The difference in properties resulted from the combined effect of the interatomic binding strength and the mixed bonding type.
{"title":"Enhanced He irradiation-resistance of M/A-site two-component MAX phase revealed via defect evolution","authors":"Yulin Wei , Xiaoyue Li , Chenhao Yang , Junxiong Liu , Ping Peng , Min Liu","doi":"10.1016/j.jnucmat.2025.155615","DOIUrl":"10.1016/j.jnucmat.2025.155615","url":null,"abstract":"<div><div>To enhance the He irradiation-resistance of the MAX phase, the irradiation-induced defect evolution of the Ti<sub>3</sub>SiC<sub>2</sub>-based two-component MAX phase has been investigated by First Principles. This clarifies the laws and mechanisms of irradiation damage in different structural two-component MAX phases, which offers a crucial reference for the screening of new nuclear structural materials. It was found that the M-site solid solution promoted the vacancies formation, inhibited interstitial dissolution of He, and reduced the damage to atomic stability caused by He-vacancies. The A-site solid solution decreased the formation energy of antisite defects and enhanced the resistance to damage induced by vacancies. He bubbles are less likely to form because both two-component MAX phases prevent He atoms from migrating and binding within the lattice and interfaces. Furthermore, the degree of lattice distortion affected the degree of property alteration and the tendency, which made the comprehensive performance of (Ti<sub>0.5</sub>Ta<sub>0.5</sub>)<sub>3</sub>SiC<sub>2</sub> and Ti<sub>3</sub>(Si<sub>0.5</sub>Al<sub>0.5</sub>)C<sub>2</sub> better. The difference in properties resulted from the combined effect of the interatomic binding strength and the mixed bonding type.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"606 ","pages":"Article 155615"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143155390","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 : 2025-02-01DOI: 10.1016/j.jnucmat.2024.155590
Hwa Jeong Han, Hyunjin Boo, Suhyeon Lee, Byung Gi Park
A chloride-based molten salt reactors (MSRs) is preferred for fast neutron energy spectrum, which offers flexibility on choosing fissile materials, but an information of material corrosion is insufficient to develop MSR. The molten salt thermal convection loop (MSTCL) has been designed and constructed from stainless steel 316L to understand material-related issues relevant to MSR. However, during testing corrosion specimens, an event of the salt leakage and heater failure was detected in the weld zone of the loop. Microstructural observation with scanning electron microscopy showed wall thinning of the external surface near the perforated hole, intergranular attack in the internal surface, and unusual crevice and crack propagation in the intact weld zone. A root cause analysis has identified three molten chloride corrosion phenomena of impurity-driven intergranular attack, stress-assisted corrosion in crevice, and high temperature oxidation and chlorination in the atmosphere. It is confirmed that stress-assisted corrosion in crevice is responsible for the salt leakage. Residual stress due to welding combined with impurity-driven intergranular attack may lead to stress-assisted corrosion in crevice. Therefore, it can be inferred that crevice and residual stress in the internal surface are major causes of wall penetration of the loop and salt leak. Since both crevice and residual stress in the weld zone may originate from the manufacturing process involving welding, sufficient consideration should be given to the design of weld joint and the welding procedure to prevent these problem-causing factors from occurring.
{"title":"Analysis of salt leakage due to corrosion-induced failure in molten salt thermal convection loop (MSTCL)","authors":"Hwa Jeong Han, Hyunjin Boo, Suhyeon Lee, Byung Gi Park","doi":"10.1016/j.jnucmat.2024.155590","DOIUrl":"10.1016/j.jnucmat.2024.155590","url":null,"abstract":"<div><div>A chloride-based molten salt reactors (MSRs) is preferred for fast neutron energy spectrum, which offers flexibility on choosing fissile materials, but an information of material corrosion is insufficient to develop MSR. The molten salt thermal convection loop (MSTCL) has been designed and constructed from stainless steel 316L to understand material-related issues relevant to MSR. However, during testing corrosion specimens, an event of the salt leakage and heater failure was detected in the weld zone of the loop. Microstructural observation with scanning electron microscopy showed wall thinning of the external surface near the perforated hole, intergranular attack in the internal surface, and unusual crevice and crack propagation in the intact weld zone. A root cause analysis has identified three molten chloride corrosion phenomena of impurity-driven intergranular attack, stress-assisted corrosion in crevice, and high temperature oxidation and chlorination in the atmosphere. It is confirmed that stress-assisted corrosion in crevice is responsible for the salt leakage. Residual stress due to welding combined with impurity-driven intergranular attack may lead to stress-assisted corrosion in crevice. Therefore, it can be inferred that crevice and residual stress in the internal surface are major causes of wall penetration of the loop and salt leak. Since both crevice and residual stress in the weld zone may originate from the manufacturing process involving welding, sufficient consideration should be given to the design of weld joint and the welding procedure to prevent these problem-causing factors from occurring.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"605 ","pages":"Article 155590"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143170549","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 : 2025-02-01DOI: 10.1016/j.jnucmat.2024.155569
Zhonghua Lu , Yanli Shi , Gaoyuan Wang , Xiuling Wang , Jianqi Qi , Tiecheng Lu
Whether the grain boundary (GB) accelerates or impedes the diffusion of hydrogen isotopes is reported with considerable inconsistency. Here the migration of tritium in (1–10) twin GB of Li4TiO4 breeder ceramic is investigated with density functional theory and kinetic Monte Carlo. The identified tritium sites along GB exhibit significantly varied segregation energies in the range of −1.57 ∼ −0.45 eV, which is attributed to the difference in the space volume around the site and the hydrogen bond. The minimum energy paths for tritium migration are obtained and the global barriers (1.12 eV and 0.88 eV along a and b, respectively) indicate a slower-than-bulk diffusion along GB. The rugged energy landscape along GB is analyzed by the kinetic and configurational components of local barriers. It's observed that the induced difference in configurational energies (i.e. the varied segregation strengths) by GB structure makes the major contribution to the raised global barrier. Diffusion coefficient for tritium migration along GB is estimated as , which is comparable to experimental data. The presented case offers insights for understanding the modified diffusion rates of segregated species by GB.
{"title":"Elevated tritium diffusion barrier from varied segregation strengths in Li4TiO4 (1–10) grain boundary","authors":"Zhonghua Lu , Yanli Shi , Gaoyuan Wang , Xiuling Wang , Jianqi Qi , Tiecheng Lu","doi":"10.1016/j.jnucmat.2024.155569","DOIUrl":"10.1016/j.jnucmat.2024.155569","url":null,"abstract":"<div><div>Whether the grain boundary (GB) accelerates or impedes the diffusion of hydrogen isotopes is reported with considerable inconsistency. Here the migration of tritium in (1–10) twin GB of Li<sub>4</sub>TiO<sub>4</sub> breeder ceramic is investigated with density functional theory and kinetic Monte Carlo. The identified tritium sites along GB exhibit significantly varied segregation energies in the range of −1.57 ∼ −0.45 eV, which is attributed to the difference in the space volume around the site and the hydrogen bond. The minimum energy paths for tritium migration are obtained and the global barriers (1.12 eV and 0.88 eV along <strong><em>a</em></strong> and <strong><em>b</em></strong>, respectively) indicate a slower-than-bulk diffusion along GB. The rugged energy landscape along GB is analyzed by the kinetic and configurational components of local barriers. It's observed that the induced difference in configurational energies (i.e. the varied segregation strengths) by GB structure makes the major contribution to the raised global barrier. Diffusion coefficient for tritium migration along GB is estimated as <span><math><mrow><mn>7</mn><mrow><mo>.</mo><mn>30</mn><mo>×</mo><mn>1</mn></mrow><msup><mrow><mn>0</mn></mrow><mrow><mo>−</mo><mn>7</mn></mrow></msup><mtext>exp</mtext><mrow><mo>(</mo><mrow><mrow><mo>−</mo><mn>0</mn></mrow><mrow><mo>.</mo><mn>98</mn><mspace></mspace><mtext>eV</mtext></mrow><mo>/</mo><msub><mi>K</mi><mi>B</mi></msub><mi>T</mi></mrow><mo>)</mo></mrow><msup><mrow><mi>m</mi></mrow><mn>2</mn></msup><mo>/</mo><mi>s</mi></mrow></math></span>, which is comparable to experimental data. The presented case offers insights for understanding the modified diffusion rates of segregated species by GB.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"605 ","pages":"Article 155569"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143170573","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 : 2025-02-01DOI: 10.1016/j.jnucmat.2024.155577
SungHoon Joung, Hyunwoo Yook, Dongju Kim, Youho Lee
The formation of the Zr-Cr eutectic (∼1320 °C), which does not occur in conventional Zirconium alloys, introduces a significant safety concern for Cr-coated Accident Tolerant Fuel (ATF). This study investigated the Peak Cladding Temperature (PCT) limit for Cr-coated ATF by examining the effects of the Zr-Cr eutectic on the mechanical integrity of Cr-coated Zr-1.1Nb claddings. To achieve this, Integral Loss of Coolant Accident (LOCA) tests and Ring Compression Tests (RCTs) were conducted on Cr-coated specimens under both steam and oxygen-free environments. The results indicated that while the formation of the eutectic phase between Zr and Cr does not result in structural failure, it reduced the ductility of the cladding. However, the impact of Zr-Cr eutectic on the reduction in ductility was overshadowed by the significant impact of the oxidation under the same conditions. The primary cause of the severe ductility loss in specimens oxidized above the eutectic onset temperature was the increased oxygen diffusion at elevated temperatures. Consequently, compared to specimens oxidized at 1204 °C, the increased oxygen concentration in the ductile layer further reduced the ductility of the cladding. Based on these findings, the pronounced reduction in ductility caused by oxidation of the Zr matrix in Cr-coated ATF cladding underscored the necessity of adhering to the current PCT limit, as long as the cladding matrix of Cr-coated ATF cladding remains Zirconium alloy. Furthermore, the excessive embrittlement observed in Zirconium alloy at temperatures above 2400 °F (1315 °C) was a key factor in establishing the current 2200 °F (1204 °C) PCT limit. As a result, extending the PCT limit beyond 1204 °C for Cr-coated ATF cladding is impractical, given the rapid oxygen diffusion and the consequent reduction in ductility at these higher temperatures. Therefore, maintaining the current 2200 °F (1204 °C) PCT limit for Cr-coated ATF cladding can serve as the effective approach for ensuring the safety of Cr-coated ATF cladding within the existing regulatory framework.
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