Journal of Nuclear Materials最新文献

{"title":"Thermochemical bounds on UCO TRISO kernel compositions","authors":"M. Poschmann,&nbsp;D. Wojtaszek,&nbsp;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_<em>x</em>. The lower bound on UC_<em>x</em> 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}

{"title":"Dispersed barrier hardening modeling on depth-distributed helium bubbles in iron-based alloys","authors":"Xinrun Chen ,&nbsp;Tatsuya Suzuki ,&nbsp;Phongsakorn Prak Tom ,&nbsp;Bo Li ,&nbsp;Zongda Yang ,&nbsp;Sho Kano ,&nbsp;Takuya Yamamoto ,&nbsp;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⁺ 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.</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}

{"title":"Understanding and removing FIB artifacts in metallic TEM samples using flash electropolishing","authors":"Danny J. Edwards ,&nbsp;Alan Schemer-Kohrn ,&nbsp;Matt Olszta ,&nbsp;Ramprashad Prabhakaran ,&nbsp;Yuanyuan Zhu ,&nbsp;Jing Wang ,&nbsp;Jacob Haag ,&nbsp;Osman El Atwani ,&nbsp;Timothy G. Lach ,&nbsp;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⁺ 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.</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}

{"title":"Effect of cold forging on the microstructure and corrosion behavior of type 316L stainless steel in molten FLiNaK salt","authors":"Jie Qiu ,&nbsp;Ryan D. Hayes ,&nbsp;Ho Lun Chan ,&nbsp;Haley Williams ,&nbsp;Digby D. Macdonald ,&nbsp;Raluca O. Scarlat ,&nbsp;Djamel Kaoumi ,&nbsp;John R. Scully ,&nbsp;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}

{"title":"Enhanced He irradiation-resistance of M/A-site two-component MAX phase revealed via defect evolution","authors":"Yulin Wei ,&nbsp;Xiaoyue Li ,&nbsp;Chenhao Yang ,&nbsp;Junxiong Liu ,&nbsp;Ping Peng ,&nbsp;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 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 (Ti_0.5Ta_0.5)₃SiC₂ and Ti₃(Si_0.5Al_0.5)C₂ 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}

{"title":"Analysis of salt leakage due to corrosion-induced failure in molten salt thermal convection loop (MSTCL)","authors":"Hwa Jeong Han,&nbsp;Hyunjin Boo,&nbsp;Suhyeon Lee,&nbsp;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}

{"title":"Elevated tritium diffusion barrier from varied segregation strengths in Li4TiO4 (1–10) grain boundary","authors":"Zhonghua Lu ,&nbsp;Yanli Shi ,&nbsp;Gaoyuan Wang ,&nbsp;Xiuling Wang ,&nbsp;Jianqi Qi ,&nbsp;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 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 <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}

{"title":"Exploring the Peak Cladding Temperature Limit of Cr-coated ATF Cladding by Assessing the Impact of the Zr-Cr Eutectic on the Structural Integrity of Cladding","authors":"SungHoon Joung,&nbsp;Hyunwoo Yook,&nbsp;Dongju Kim,&nbsp;Youho Lee","doi":"10.1016/j.jnucmat.2024.155577","DOIUrl":"10.1016/j.jnucmat.2024.155577","url":null,"abstract":"<div><div>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.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"605 ","pages":"Article 155577"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143171595","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}

{"title":"Abnormal growth of loops by repulsive interaction between loops and dislocations","authors":"Shen Li ,&nbsp;Yipeng Li ,&nbsp;Ziqi Cao ,&nbsp;Yifan Ding ,&nbsp;Xiaoyong Wu ,&nbsp;Ruiqian Zhang ,&nbsp;Guang Ran","doi":"10.1016/j.jnucmat.2024.155589","DOIUrl":"10.1016/j.jnucmat.2024.155589","url":null,"abstract":"<div><div>Pre-existing dislocations have a significant impact on the evolution of radiation defects, which in turn affect the mechanical properties of materials after irradiation. However, systematic research into the question of how pre-existing dislocations affect the evolution of defects is essential and has yet to be carried out. In this work, the effect of pre-existing dislocations on loop evolution in FeCrAl alloys is investigated by <em>in-situ</em> transmission electron microscopy (TEM) using Fe⁺+He⁺+H₂⁺ tri-beam irradiation. It is found that the dislocations exhibit a strong sink effect and the values of the number density and size of the dislocation loops are lower in regions with pre-existing dislocations. As the calculation of sink strength shows, the sink effect of dislocations can be weakened due to the initiation of dislocation gliding, by the absorption of point defects and by climbing. And this impaired sink effect is the main reasons for the formation of a V-shaped density profile. A coexisting repulsive interaction between loops and dislocations has been proposed, which also has an important effect on the evolution morphology of loops, and its validity has been demonstrated by the interaction energy calculations. The repulsive interaction of dislocations pushes the defects away, leading to a higher probability of nucleation and coalescence. This can lead to a greater number and abnormally rapid growth of loops. In addition, a stronger surface effect is found for the thinner foil, where the density is lower and the size is smaller. The finger-shaped loops are found to be absent in the thinner samples due to the annihilation of the large loops.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"605 ","pages":"Article 155589"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143171599","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}

{"title":"Mechanistic nuclear fuel performance modeling of uranium nitride","authors":"Jason T. Rizk ,&nbsp;Michael W.D. Cooper ,&nbsp;Pierre-Clément A. Simon ,&nbsp;Anton J. Schneider ,&nbsp;David A. Andersson ,&nbsp;Stephen R. Novascone ,&nbsp;Christopher Matthews","doi":"10.1016/j.jnucmat.2024.155604","DOIUrl":"10.1016/j.jnucmat.2024.155604","url":null,"abstract":"<div><div>Uranium mononitride (UN) is a nuclear fuel candidate for advanced reactor designs and an alternative being considered for light water reactors due to its higher thermal conductivity and uranium density than UO₂. As with any nuclear fuel, swelling and fission gas release are important factors for safety, while also being some of the hardest phenomena to predict with a high degree of confidence. Getting a grasp on the gas swelling behavior and release is crucial to lower the barrier for UN utilization. An accelerated swelling rate at high temperatures observed experimentally, sometimes referred to as “breakaway swelling,” further complicates the prediction of fuel performance of UN. A mechanistic model has been developed using a multiscale approach to describe the intragranular and intergranular fission gas behavior. Lower-length-scale calculations have been employed to inform models of the gas and self-diffusion behavior, resolution rate, and bubble shape. Leveraging previous work on high burnup UO₂, two populations of intragranular bubbles are considered; small bulk bubbles and larger bubbles located along dislocations. The dislocation bubbles were found to be crucial to the overall swelling behavior, and the breakaway swelling transition was associated with the transition in the gas atom diffusion mechanism from an irradiation-induced athermal diffusion regime at lower temperatures to an intrinsic thermal equilibrium regime at higher temperatures, accelerating the growth of the dislocation bubbles. Similarly, the threshold for fission gas release was associated with the grain boundary vacancy diffusivity surpassing the gas atom diffusivity at sufficiently high temperatures, allowing the over-pressurized grain boundary bubble to grow in size and interconnect. Using thermo-mechanical models with the fission gas model, two integral fuel pin assessment cases were simulated. This work demonstrates the ability of a multiscale approach to accelerate the understanding of advanced fuel forms when experimental data is limited.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"606 ","pages":"Article 155604"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143155302","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}