Journal of Nuclear Materials最新文献

{"title":"Corrosion and mechanical behavior of novel alumina forming steels in molten lead","authors":"Facundo Masari ,&nbsp;Peter Szakalos ,&nbsp;Christopher Petersson ,&nbsp;José M. Torralba ,&nbsp;Mónica Campos","doi":"10.1016/j.jnucmat.2024.155587","DOIUrl":"10.1016/j.jnucmat.2024.155587","url":null,"abstract":"<div><div>Three new multi-phase alumina-forming steels with compositions Fe-(10–14.5)Cr-(10–12)Ni-3.5Al (wt.%) were exposed to stagnant lead at 550 and 650 °C for up to 1000 h The experimental alloys formed stable and protective alumina (Al₂O₃) layers at both temperatures, crucial for preventing lead penetration and material degradation. In contrast, 316 L and T91 steels, candidate materials for nuclear applications, showed significant oxidation and lead penetration, particularly at the higher temperature. The designed alloys retained their mechanical properties after exposure, with one of them even increasing yield strength due to phase transformations. The findings highlight the potential of these new alloys with no reactive elements and no thermomechanical treatments, to operate in environments with high-temperature liquid lead, such as Gen IV nuclear reactors or high-temperature concentrated solar power plants.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"605 ","pages":"Article 155587"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143170550","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":"A new model of fission gas bubble growth and mechanism analysis for U-xZr fuels","authors":"Xiaoxiao Mao ,&nbsp;Xingdi Chen ,&nbsp;Xiaobin Jian ,&nbsp;Feng Yan ,&nbsp;Shurong Ding","doi":"10.1016/j.jnucmat.2024.155578","DOIUrl":"10.1016/j.jnucmat.2024.155578","url":null,"abstract":"<div><div>U-<em>x</em>Zr alloys have a promising application prospect in advanced nuclear fuel elements, and their macroscale volume growth under the extreme service environments are attracting more attention. In this study, innovative volume growth modeling and mechanism analysis are performed for various U-<em>x</em>Zr alloys under different operation conditions. Specially, based on the creep test results in the references, the macroscale thermal creep models are newly developed for solid U-<em>x</em>Zr alloys within a temperature range, implicitly reflecting the effects of phase fraction; for the bubble contained region of equivalent spherical fuel grain, the established thermal creep models are involved in the mechanical constitutive relations for the solid fuel skeleton; the finite element equations are derived for the displacement fields of bubble contained region and numerically implemented, obtaining the multi-level variables of macroscale volume growth, the local porosity and the average porosity. The predictions of irradiation swelling for different U-<em>x</em>Zr alloys agree well with the experimental data at 743 K or 903 K; the fast-swelling phenomena due to various thermal creep contributions could be captured, demonstrating the progressiveness of the developed new models and algorithms. The numerical simulation results indicate that: (1) under the irradiation temperature of 603 K or 703 K, dislocation creep mechanism of fuel skeleton is dominated, due to higher internal and external pressure differences; (2) at the high temperatures of 803 K and 903 K, the thermal diffusion creep deformations of fuel skeleton contribute dominantly to the macroscale volume growth of U-<em>x</em>Zr alloys over the whole irradiation process; (3) under zero external pressure the sharp increase phenomena of fission gas swelling become more and more distinct with the rise of irradiation temperature, stemming from the quickened diffusion of fission gas atom and the enhanced creep deformations of fuel skeleton; at 903 K the fuel skeleton is prone to creep deformation, leading to significant inhibition of bubble growth by a small external pressure. This research provides important theoretical models and algorithms for simulation of the irradiation-induced thermo-mechanical behaviors in U-<em>x</em>Zr-based fuel elements or assemblies.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"605 ","pages":"Article 155578"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143170572","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":"Molecular dynamics simulations of interaction between a super edge dislocation and interstitial dislocation loops in irradiated L12-Ni3Al","authors":"Cheng Chen ,&nbsp;Dongyang Qin ,&nbsp;Yiding Wang ,&nbsp;Fei Xu ,&nbsp;Jun Song","doi":"10.1016/j.jnucmat.2024.155541","DOIUrl":"10.1016/j.jnucmat.2024.155541","url":null,"abstract":"<div><div>The study employed MD simulations to investigate the interactions between a <span><math><mrow><mo>〈</mo><mrow><mn>1</mn><mover><mn>1</mn><mo>¯</mo></mover><mn>0</mn></mrow><mo>〉</mo></mrow></math></span> super-edge dislocation, consisting of the four Shockley partials, and interstitial dislocation loops (IDLs) in irradiated <span><math><mrow><mi>L</mi><msub><mn>1</mn><mn>2</mn></msub></mrow></math></span>-Ni₃Al. Accounting for symmetry breakage in the <span><math><mrow><mi>L</mi><msub><mn>1</mn><mn>2</mn></msub></mrow></math></span> lattice, the superlattice planar faults with four distinct fault vectors have been considered for different IDL configurations. The detailed dislocation reactions and structural evolution events were identified as the four partials interacted with various IDL configurations. The slipping characteristics of Shockley partials within the IDLs and the resultant shearing and looping mechanisms were also clarified, revealing distinct energetic transition states determined by the fault vectors after the Shockley partials sweeping the IDL. Furthermore, significant variations in critical resolved shear stress (CRSS) required for the super-edge dislocation to move past the IDL were observed, attributed to various sizes and faulted vectors of enclosed superlattice planar faults in the IDLs. The current study extends the existing dislocation-IDL interaction theory from pristine FCC to <span><math><mrow><mi>L</mi><msub><mn>1</mn><mn>2</mn></msub></mrow></math></span> lattice, advances the understanding of irradiation hardening effects in <span><math><mrow><mi>L</mi><msub><mn>1</mn><mn>2</mn></msub></mrow></math></span>-<span><math><mrow><mi>N</mi><msub><mi>i</mi><mn>3</mn></msub></mrow></math></span>Al, and suggests potential applicability to other <span><math><mrow><mi>L</mi><msub><mn>1</mn><mn>2</mn></msub></mrow></math></span> systems.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"605 ","pages":"Article 155541"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143171148","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":"Accelerated prediction of lattice thermal conductivity of Zirconium and its alloys: A machine learning potential method","authors":"Fan Yang ,&nbsp;Di Wang ,&nbsp;Jiaxuan Si ,&nbsp;Jianqiao Yu ,&nbsp;Zhen Xie ,&nbsp;Xiaoyong Wu ,&nbsp;Yuexia Wang","doi":"10.1016/j.jnucmat.2024.155603","DOIUrl":"10.1016/j.jnucmat.2024.155603","url":null,"abstract":"<div><div>Zirconium alloy coating is an important direction for the modification of nuclear cladding materials. Thermal conductivity is a critical property of cladding materials. With extensively studying phonon-electron non-equilibrium energy transfer processes in the thermal transport of zirconium alloy coating, to distinguish the contributions from phonon and electron thermal conductivity of Zr alloys becomes crucial and necessary. In this work, we successfully predicted the lattice thermal conductivities of zirconium, Zr-Sn and Zr-Nb using machine learning potentials. Sn and Nb doping leads to a significant decrease in lattice thermal conductivity, which is mainly due to the alterations in phonon group velocity and phonon scattering. The larger atomic mass of doping elements and weakened interatomic interactions of Zr-Nb together lead to a significant decrease in phonon group velocity. Doping Sn and Nb also increases phonon-phonon scattering rate and three-phonon scattering channels, resulting in a shortening in phonon lifetime and a decrease in lattice thermal conductivity.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"605 ","pages":"Article 155603"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143171594","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 stability and mechanical property of novel high-Si high-Cr reduced activation ferritic/martensitic steels at high temperatures","authors":"Qianfu Pan ,&nbsp;Sen Ge ,&nbsp;Chao Sun ,&nbsp;Gaixia Wang ,&nbsp;Yu Wu ,&nbsp;Xiaoe Xu ,&nbsp;Huiqun Liu","doi":"10.1016/j.jnucmat.2024.155602","DOIUrl":"10.1016/j.jnucmat.2024.155602","url":null,"abstract":"<div><div>The present work investigated the microstructural stability and mechanical property of four novel high-Si and high Cr reduced activation ferritic/martensitic steels at elevated temperature. Alloy plate samples were normalized at 1373 K for 1 h, tempered at 1023 K for 1 h, and then aged at 873 K for 1000, 2000, and 3000 h In the tempered state, M₂₃C₆ precipitates were distributed along grain and lath boundaries, while MX precipitates were uniformly dispersed in the matrix containing different amounts of ferrites, which was similar to with the calculated result. The microstructure of the designed alloys exhibited high-thermal stability even after 3000 h aging, with the martensitic grain size and ferrite content nearly unchanged. However, M₂₃C₆ were coarsened with increasing the aging time. Additionally, with increasing W content, the coarsening rate significantly decreased. After aging for 1000 h, the designed alloys precipitated needle-like Laves phases with a faster coarsening rate, and the size and volume fraction increased with W and Si content. While VN precipitates exhibited significantly higher stability, maintaining a constant particle size (60 ∼ 80 nm) even after aging for 3000 h, which is attributed to variations in the diffusion coefficients of elements. The designed alloys exhibited high yield strength (488 ∼ 548 MPa at room temperature) in the 3000 h-aged state, surpassing that of commercial EP823 (462 MPa), where the strengthening mechanisms were also discussed.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"606 ","pages":"Article 155602"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143154914","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":"Atomistic investigation of the evolution of He cluster-loop complex in fcc nickel","authors":"Ping Yu,&nbsp;Ali Wen,&nbsp;Zhenbo Zhu,&nbsp;Cuilan Ren,&nbsp;Hefei Huang","doi":"10.1016/j.jnucmat.2025.155632","DOIUrl":"10.1016/j.jnucmat.2025.155632","url":null,"abstract":"<div><div>The evolution of He bubble-loop complexes in fcc Ni is studied using atomistic simulations. The influences of temperature, He concentration and formation sequence (He bubbles formation cooperatively with or after a faulted dislocation loop) on the evolutions are systematically calculated and analyzed. Results show He bubbles nucleate at the edges of dislocation loops rather than inside them in all conditions, leading to a reduction of the formation energy of loops. We reveal two growth processes of the complexes, depending primarily on the formation sequence. When bubbles and loops grow cooperatively, the loop lines can bend around to enclose bubbles, creating complexes containing bubbles both at the loop edge and within the loop plane. When a bubble forms after a dislocation loop creation, the resulting complex contains bubbles only at the loop edge. The size of the bubble-loop complex increases with He concentration owing to additional SIAs kicked out by the He bubble, irrespective of its formation sequence. This study provides atomistic insights into He-enhanced nucleation and growth of dislocation loops in fcc metals and will help in developing a better understanding of the microstructure evolution in irradiated Ni materials.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"606 ","pages":"Article 155632"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143154919","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":"Importance of basal slip on the plastic deformation behavior of zirconium alloy Zr702","authors":"X. Chen ,&nbsp;A. Chapuis ,&nbsp;W.J. He ,&nbsp;Q. Liu","doi":"10.1016/j.jnucmat.2025.155655","DOIUrl":"10.1016/j.jnucmat.2025.155655","url":null,"abstract":"<div><div>The plastic deformation behavior of a rolled Zr702 zirconium plate is investigated experimentally and with visco-plastic self-consistent (VPSC) simulations. Compression tests along the RD, TD and ND (rolling, transverse and normal directions, respectively) are done at room temperature and the plastic anisotropy (<em>r</em>-values or Lankford coefficients) is measured. EBSD measurements are used to assess the initial texture and the twin volume fraction after 10 % strain compression. Compression stress-strain curves are used to fit the critical resolved shear stress (CRSS) for prismatic, basal and 〈<em>c</em> + <em>a</em>〉 slip, and the CRSS for {10–12} tension twinning is fitted to reproduce the measured twin volume fraction. We found that basal slip is necessary to reproduce the plastic flow anisotropy, but stress strain curves can be reproduced with several sets of material parameters. Two set of material parameters are presented: first with the CRSS for basal slip equal to the CRSS for pyramidal 〈<em>c</em> + <em>a</em>〉 slip, second with the CRSS for basal slip half that for 〈<em>c</em> + <em>a</em>〉 slip. The predicted plastic strain anisotropy (<em>r</em>-value) strongly depends on the CRSS for basal slip. Simulations proved that the predicted <em>r</em>-values match the experimental ones only when the CRSS for basal slip is half that of 〈<em>c</em> + <em>a</em>〉 slip. Another self-consistent model is used to confirm that the CRSS for basal slip is half that of 〈<em>c</em> + <em>a</em>〉 slip. Slip traces analysis confirmed the activation of basal slip and pyramidal 〈<em>c</em> + <em>a</em>〉 slip, and pyramidal 〈a〉 slip was possibly activated. The different material parameters are used to predict the texture after plane strain deformation and simulations show that the textures are quite similar.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"606 ","pages":"Article 155655"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143155317","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 He ion irradiation on the early corrosion behaviour of SIMP steel in liquid lead-bismuth eutectic","authors":"Zhiwei Ma ,&nbsp;Peng Jin ,&nbsp;Xing Gao ,&nbsp;Lilong Pang ,&nbsp;Tielong Shen ,&nbsp;Zhiguang Wang","doi":"10.1016/j.jnucmat.2025.155623","DOIUrl":"10.1016/j.jnucmat.2025.155623","url":null,"abstract":"<div><div>He-ion irradiated and pristine SIMP steel samples were corroded in oxygen-saturated liquid LBE at 550°C for 24 h. The results show that the oxide layer on the irradiated sample is significantly thicker than that on the pristine sample. In the irradiated sample, the inner oxide layer is notably enriched with Si and Cr at the martensitic lath boundaries, causing severe depletion of these elements within the laths. This significantly accelerates the oxidation of the irradiated sample in liquid LBE in combination with the increased surface roughness of the irradiated sample, the radiation enhanced elemental diffusion and the voids in the inner oxide layer.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"606 ","pages":"Article 155623"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143155319","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":"Ab initio study of helium in titanium beryllides","authors":"D.V. Bachurin,&nbsp;C. Stihl,&nbsp;P.V. Vladimirov","doi":"10.1016/j.jnucmat.2025.155646","DOIUrl":"10.1016/j.jnucmat.2025.155646","url":null,"abstract":"<div><div>Be₁₂Ti compound is proposed as a neutron multiplier for tritium-breeding blankets in the demonstration fusion reactor DEMO. Recent experimental studies suggested that Be₁₂Ti could contain additions of other phases such as Be₂Ti and Be₁₇Ti₂. In light of these findings, investigation of helium behavior and its binding with vacancy traps in the crystal lattices of these phases is crucial. The paper employs <em>ab initio</em> methods to calculate the helium binding energy with various monovacancy types, as well as the helium solution energies at interstitial sites. The solution energy of helium in all non-equivalent interstitial sites of the titanium beryllides is at least 0.6 eV lower than that for pure beryllium. In the titanium beryllides, helium exhibits stronger binding with the titanium vacancy than with the beryllium vacancy. The binding energy of helium to a vacancy in both Be₁₂Ti and Be₁₇Ti₂ is almost the same as in pure beryllium, except for Be₂Ti, which has a lower binding energy. When helium is in the vicinity of a vacancy, it causes the displacement of adjacent beryllium atom into the initial vacancy, while helium substitutes the displaced beryllium atom. Some helium atoms may become trapped by a vacancy being outside of it. The obtained results are crucial for the future assessment of interstitial helium diffusion and helium bubble nucleation in titanium beryllides.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"606 ","pages":"Article 155646"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143155508","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":"In-situ strain behavior and BISON simulations of Zircaloy cladding subjected to temperature cycling separate-effects tests in a steam environment","authors":"Jennifer I. Espersen ,&nbsp;Nathan A. Capps ,&nbsp;Mackenzie J. Ridley ,&nbsp;Sam B. Bell ,&nbsp;Nicholas R. Brown","doi":"10.1016/j.jnucmat.2024.155570","DOIUrl":"10.1016/j.jnucmat.2024.155570","url":null,"abstract":"<div><div>Understanding fuel system performance during anticipated transients without scram (ATWSs) in boiling water reactors (BWRs) is necessary for refining current and future safety limits. High-fidelity material models and simulations are fundamental to rigorous assessment of zirconium-based cladding performance. However, experimental thermomechanical data during simulated ATWSs to validate these modes are limited. To provide relevant in-situ data, Zircaloy-4 cladding was subjected to cyclic heating in a steam environment to simulate an out-of-pile BWR ATWS. Digital image correlation was used to capture the cladding strain behavior in-situ for comparison against simulations using the BISON finite element code. Conventional high-temperature models were compared using multiple schemes to gain a better understanding of the applicability of three BISON models to BWR ATWS: (1) the default combination of creep models in BISON, (2) the high-temperature Erbacher model alone, and (3) the low-temperature Limback-Andersson model alone. The cases run with the Limback-Andersson model alone produced the lowest root mean square error (RMSE). The lowest RMSE for the Limback-Andersson model alone was 0.659%, and the highest RMSE reported was 4.22%. A data gap within the model in the temperature regime of interest was also identified, and to account for this gap, the current model in BISON is linearly interpolated between two separate datasets. This evaluation highlights the need to either develop a new model or to improve the existing model to capture transient creep effects resulting from a cyclic temperature transient.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"605 ","pages":"Article 155570"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143170555","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}