Journal of Nuclear Materials最新文献

{"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}

{"title":"Microstructural evolution in ion irradiated cold spray Cr coated Zr-alloy","authors":"Tyler Dabney ,&nbsp;K.N. Sasidhar ,&nbsp;Evan Willing ,&nbsp;Carson Lukas ,&nbsp;Kyle Quillin ,&nbsp;Hwasung Yeom ,&nbsp;Kumar Sridharan","doi":"10.1016/j.jnucmat.2025.155652","DOIUrl":"10.1016/j.jnucmat.2025.155652","url":null,"abstract":"<div><div>Chromium coating deposited on zirconium-alloy using cold spray technology is being considered as a potential approach for increasing the accident tolerance of light water reactors. In this study, high energy Xe²⁶⁺ ion irradiation was performed on cold spray coated Cr/Zircaloy-4 system at 350 °C to investigate structural changes within the bulk Cr coating and at the Cr/Zircaloy-4 interface. High resolution structural and compositional analysis of the as-deposited coating revealed crystallographic coherence at the interface and nanometric segregation of Cr and Fe on the Zr-alloy side of the interface. Irradiation led to homogenization of Fe and the formation of a ∼20 nm thick amorphous layer at the Cr/Zircaloy-4 interface. Thermodynamic calculations revealed that both Fe dissolution and amorphization are favored due to the increased enthalpy provided by the high energy irradiation. The coatings exhibited a heterogeneous structure over multiple length scales, consisting of elongated and dynamically recrystallized ultrafine or nanocrystalline grains, high dislocation density, gradation in grain size, and interparticle boundaries. After irradiation, the interparticle boundaries were obliterated to be replaced by recrystallized grains and subtle grain growth. Consistent with these structural changes, nanoindentation testing exhibited slight softening of the coating after irradiation.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"606 ","pages":"Article 155652"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143155395","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":"Neutron flux impact on rate of expansion of quartz","authors":"Ippei Maruyama ,&nbsp;Kenta Murakami ,&nbsp;Takahiro Ohkubo ,&nbsp;Shohei Sawada ,&nbsp;Osamu Kontani ,&nbsp;Takafumi Igari ,&nbsp;Masaki Kawai ,&nbsp;Junji Etoh","doi":"10.1016/j.jnucmat.2025.155631","DOIUrl":"10.1016/j.jnucmat.2025.155631","url":null,"abstract":"<div><div>The radiation-induced expansion of concrete aggregate is a significant concern in the aging management of concrete biological shielding in nuclear power plants. Understanding the sensitivity of rock-forming minerals, particularly quartz, to neutron radiation is essential in this context. In this study, we investigated the neutron irradiation effects on different types of quartz, including synthetic quartz, metachert, sandstone, and granodiorite, under irradiation temperatures ranging from 45 to 62 ℃ and the displacement damage in quartz ranging from 0.01 to 0.23 dpa. The rate of irradiation-induced expansion was determined using X-ray diffraction/Rietveld analysis, revealing a neutron flux dependency. Our findings suggest that radiation-induced relaxation or healing processes occur within quartz. To explain the observed flux-dependent and temperature-dependent radiation-induced expansion of quartz, we propose a two-phase model that considers the pristine and expanded phases. This model, based on four key variables, i.e. neutron flux, the equivalent cross-sections for phase change, and healing parameter, successfully reproduces the radiation-induced expansion behaviors observed in quartz. The model further indicates that radiation-induced relaxation, potentially linked to the diffusion of silicon (Si) or oxygen (O) within the quartz grain, plays a healing role that mitigates radiation-induced volume expansion.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"606 ","pages":"Article 155631"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143349353","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":"Synergistic effect of simultaneous proton irradiation and molten lead-bismuth eutectic corrosion on FeCrAlTi(Mo) coatings","authors":"Jian Yang ,&nbsp;Wei Zhang ,&nbsp;Jijun Yang","doi":"10.1016/j.jnucmat.2024.155600","DOIUrl":"10.1016/j.jnucmat.2024.155600","url":null,"abstract":"<div><div>The synergistic effect of simultaneous proton irradiation (with a total fluence of 2.0 × 10¹⁷ ions/cm²) and lead-bismuth eutectic (LBE) corrosion (at 550 °C for 46 h) on FeCrAlTi(Mo) coatings was investigated. The surface morphology, composition distribution, crystal structure of the formed oxide scale, and structural evolution of the two coatings were systematically analyzed using SEM and TEM. The corrosion results revealed the accelerated LBE corrosion behaviour of the proton-irradiated FeCrAlTi coating, which could be attributed to irradiation-induced defects leading to enhanced diffusion and irradiation-induced grain growth, resulting in the rapid formation of a protective oxide scale. However, proton irradiation decelerated the corrosion rate of the FeCrAlTiMo coating in high-temperature LBE, which could be attributed to the irradiation-induced annihilation of the free volume, which slowed down the diffusion dynamics. Whether proton irradiation accelerated or decelerated corrosion depended on the crystal structures of the FeCrAlTi and FeCrAlTiMo coatings during the corrosion process. Moreover, the novel mechanism of irradiation-decelerated corrosion provides a new concept for designing amorphous nuclear structural materials with improved synergistic corrosion and irradiation resistance.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"605 ","pages":"Article 155600"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143170531","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":"A peridynamic model for oxidation of T91 steel in liquid lead-bismuth eutectic","authors":"Chenwen Tian ,&nbsp;Zhikun Zhou ,&nbsp;Juan Du ,&nbsp;Shuaiqi Fan ,&nbsp;Ziguang Chen","doi":"10.1016/j.jnucmat.2024.155594","DOIUrl":"10.1016/j.jnucmat.2024.155594","url":null,"abstract":"<div><div>In this paper, we present a reaction-diffusion peridynamic model for the oxidation of T91 steel in liquid lead-bismuth eutectic (LBE). By integrating oxidation kinetics with diffusion dynamics, our model is applicable to a broad range of high-temperature oxidation processes. Validation against experimental data of oxide scale growth for T91 steel in oxygen-saturated LBE at 400–550°C and up to 13,000 h confirms its accuracy. Furthermore, we demonstrate the model's enhanced accuracy relative to traditional formula-based kinetic models. The new framework is utilized to investigate the influence of temperature and oxygen concentration on oxide scale evolution, indicating that elevated oxygen concentrations and temperatures significantly accelerate oxidation. Additionally, we explore the oxidation behavior of T91 steel with a non-uniform surface and oxide scale cracks, revealing that the outer magnetite oxide scale grows more rapidly than the inner Fe-Cr spinel scale in concave regions. Our model captures the transition of the oxide scale from an irregular to a flat, uniform morphology, aligning with experimental observations. This model demonstrates considerable potential for simulating high-temperature oxidation in complex geometries, including irregular surfaces and cracks, across various environmental conditions.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"605 ","pages":"Article 155594"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143170551","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}