Nuclear Engineering and Design最新文献

{"title":"Near-wall void distribution characterization in pebble bed reactor using gamma-ray CT and DEM simulation","authors":"Ahmed Jasim ,&nbsp;Mauricio Maestri ,&nbsp;Abdullah Al Zubaidi ,&nbsp;Omar Farid ,&nbsp;Muthanna Al-Dahhan","doi":"10.1016/j.nucengdes.2026.114813","DOIUrl":"10.1016/j.nucengdes.2026.114813","url":null,"abstract":"<div><div>Accurate characterization of void-fraction distributions in pebble-bed reactors (PBRs) is essential for predicting flow, heat transfer, and neutronic behavior. High-fidelity experimental benchmark data for validating such predictions remain scarce, largely due to the challenges of non-invasive measurements. In this study, gamma-ray computed tomography (CT) was employed to measure radial and cross-sectional porosity in a laboratory-scale pebble bed containing <span><math><mrow><mn>6</mn><mspace></mspace><mi>cm</mi></mrow></math></span> graphite pebbles. A Discrete Element Method (DEM) simulation was implemented and validated against these measurements, then applied to the full-scale geometry of the Xe-100 high-temperature gas-cooled pebble-bed reactor. Analyses included radial and axial void-fraction profiles in the cylindrical section and conical base, with particular attention to near-wall oscillations at multiple axial levels. Both axially averaged profiles, integrating over extended bed sections, and locally resolved profiles, capturing fine-scale oscillations, were evaluated. Additional analyses examined cross-sectional void distributions and the effect of pebble recirculation. The DEM results reproduced the expected near-wall oscillatory layering with a characteristic wavelength of ∼1 <span><math><mrow><mi>dp</mi></mrow></math></span> and bulk void fractions near <span><math><mrow><mn>0.40</mn></mrow></math></span> and further showed that oscillatory patterns persist into the conical region, where the first trough shifts outward, and a broader near-wall gap develops. Recirculation studies, corresponding to <span><math><mrow><mn>5</mn><mo>,</mo><mn>10</mn><mo>,</mo></mrow></math></span> and <span><math><mrow><mn>15</mn></mrow></math></span> complete bed inventory cycles, showed that structural rearrangements occur mainly during the initial passes, after which the bed attains a quasi-steady configuration. Recirculation intensified near-wall oscillations, particularly in the lower regions, but had negligible impact on bulk porosity in the cylindrical section. In the cone, however, the void fraction was elevated during dynamic operation due to pebble drainage and upward void propagation. The findings support improved neutronic and thermal-hydraulic modeling and contribute to the design and safety assessment of next-generation pebble-bed systems.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"450 ","pages":"Article 114813"},"PeriodicalIF":2.1,"publicationDate":"2026-02-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146190258","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"On the reactor physics implications of the use of UB2 as an additive within UN","authors":"Olga Negri,&nbsp;Tim Abram,&nbsp;Joel Turner","doi":"10.1016/j.nucengdes.2026.114800","DOIUrl":"10.1016/j.nucengdes.2026.114800","url":null,"abstract":"<div><div>The addition of UB₂ to UN fuel raises the onset temperature of its reaction with high-temperature steam. Natural boron contains 19.9% B-10, a strong neutron absorber that decays to Li-7 and He-4 upon neutron capture. This study explores the reactor physics implications of UN-UB₂ (10% UB₂) fuel with varying B-10 concentrations. Depletion analysis of an infinite lattice has shown that UN-UB₂ with 0–2% B-10 maintained cycle lengths similar to UN at equivalent U-235 enrichment and exhibited reduced reactivity swings. At 1500 effective full power days (EFPD), noble gas production in UN-UB₂ (1% B-10) was comparable to UO₂, with helium levels similar to krypton. The fuel temperature coefficient (FTC) was consistently negative across all materials, depletion levels, and temperatures. These results indicate that UN-UB₂ (1% B-10) offers comparable performance to UO₂ with benefits in enrichment, thermal conductivity, manageable helium generation, and effective reactivity control for LWRs.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"450 ","pages":"Article 114800"},"PeriodicalIF":2.1,"publicationDate":"2026-02-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146190257","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Pre-conceptual neutronic design and feasibility analysis of a thorium-based lead‑bismuth cooled small modular reactor","authors":"Xianbao Yuan ,&nbsp;Changxiao Guan ,&nbsp;Binhang Zhang ,&nbsp;Yonghong Zhang","doi":"10.1016/j.nucengdes.2026.114815","DOIUrl":"10.1016/j.nucengdes.2026.114815","url":null,"abstract":"<div><div>Small Modular Reactors (SMRs) are regarded as an important direction for future nuclear energy development due to their enhanced safety, flexible deployment capability and lower investment costs. In this study, neutronics simulations were carried out using OpenMC to design and analyze a long-life and inherently safe thorium-based lead‑bismuth cooled small modular reactor (TL-SMR). The core employs a thorium‑uranium fuel cycle and is cooled by liquid lead‑bismuth eutectic (LBE), enabling low-pressure operation and passive heat dissipation. The core can operate for more than 20 effective full-power years without refueling, achieving deep burnup. The key core parameters were analyzed, including control rod bank worths, radial and axial power factors, effective delayed neutron fractions and reactivity coefficients. The results indicate that the reactor exhibits negative reactivity coefficients, which ensure self-regulation under transient conditions. Throughout the operational cycle, the power distribution remains uniform and reasonable with sufficient shutdown margin. In summary, the TL-SMR demonstrates both long operational lifetime and favorable inherent safety characteristics. This study provides strong evidence for the technical feasibility of such reactors and offers valuable support for the further development of advanced reactor systems with inherent safety features.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"450 ","pages":"Article 114815"},"PeriodicalIF":2.1,"publicationDate":"2026-02-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146190731","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"PIV study on subchannel cross-flow characteristics in 19-pin wire-wrapped bundle channels","authors":"Rongjie Li ,&nbsp;Wukun Zhu ,&nbsp;Chengwen Qiang ,&nbsp;Minghan He ,&nbsp;Han Wang ,&nbsp;Lu Liu ,&nbsp;Sicheng Wang ,&nbsp;Dajun Fan","doi":"10.1016/j.nucengdes.2026.114810","DOIUrl":"10.1016/j.nucengdes.2026.114810","url":null,"abstract":"<div><div>The China Initiative Accelerator Driven System (CiADS) holds significant promise in the transmutation of long-lived nuclear waste and the efficient utilization of nuclear fuel. Understanding the coolant flow characteristics within reactor fuel assemblies is crucial for ensuring the safe and economical operation of nuclear facilities. In this study, a visual hydraulic experimental platform was constructed to investigate the flow velocity distribution within the fuel assembly of CiADS. Utilizing particle image velocimetry technology, the transverse velocity distribution at subchannel interfaces within a 19-pin wire-wrapped fuel assembly was systematically investigated. The results reveal a distinct cosine-type distribution of the normalized transverse velocity at the interfaces between internal subchannels, with key flow features demonstrating robust Reynolds-number independence. Notably, significant wall-induced asymmetries alter this distribution at the interface between the internal subchannel and the side-wall subchannel. A novel transverse mixing model is developed, which shows close agreement with experimental data, confirming its predictive capability for inter-subchannel mixing in wire-wrapped bundles. This research further confirms the periodic disturbance characteristics induced by wire-wraps on transverse flow and provides a theoretical basis for predicting normalized transverse velocity in wire-wrapped bundles of different sizes, thereby deepening the understanding of the transverse mixing mechanism within rod bundles.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"450 ","pages":"Article 114810"},"PeriodicalIF":2.1,"publicationDate":"2026-02-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146190730","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Comparative analysis of machine learning methods for correcting uranium loading measurement errors in pebble-bed HTGR spherical fuel elements","authors":"Yanlong Wen,&nbsp;Hongjian Zhang,&nbsp;Haiyan Xiao,&nbsp;Qing Zhu,&nbsp;Liguo Zhang,&nbsp;Hong Li","doi":"10.1016/j.nucengdes.2026.114807","DOIUrl":"10.1016/j.nucengdes.2026.114807","url":null,"abstract":"<div><div>Accurate determination of uranium loading in spherical fuel elements of high-temperature gas-cooled reactors (HTGRs) is essential for nuclear material accounting and reactor safety. Conventional γ-ray spectrometry based on a single characteristic peak (185.7 keV) suffers from significant variability due to differences in fuel sphere dimensions and TRISO particle distributions, yielding a coefficient of variation (CV) of around 1%.</div><div>In this study, a machine learning–based error correction framework was developed to enhance UO₂ loading measurements. Three representative algorithms—Random Forest (RF), Support Vector Regression (SVR), and Multilayer Perceptron (MLP)—were trained on Geant4-simulated γ-ray data, incorporating both raw counts and composite features derived from six characteristic γ lines. Results show that all models reduced the CV to 0.3–0.4%, with the MLP achieving the best performance (CV ≤ 0.35%).</div><div>Further comparative analysis indicated that the three algorithms emphasize different mathematical solution spaces: RF captures discrete threshold effects, SVR provides smooth nonlinear fitting via kernel methods, and MLP flexibly spans a richer function space to approximate the mixed discrete–continuous nature of γ-ray attenuation. The integration of these perspectives demonstrates not only the potential of machine learning for nuclear material measurement but also the complementarity of different models in balancing accuracy, robustness, and interpretability.</div><div>This work provides a practical framework for improving non-destructive UO₂ loading measurements in HTGR fuel elements, supporting more reliable nuclear material accounting and contributing to advanced reactor fuel management.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"450 ","pages":"Article 114807"},"PeriodicalIF":2.1,"publicationDate":"2026-02-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146190734","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"CFD simulation methodology for large-scale heavy liquid metal facility relevant for MYRRHA: Application to E-SCAPE facility","authors":"S. Lopes ,&nbsp;S. Keijers ,&nbsp;L. Koloszar ,&nbsp;J. Pacio ,&nbsp;P. Planquart ,&nbsp;K. Van Tichelen","doi":"10.1016/j.nucengdes.2026.114798","DOIUrl":"10.1016/j.nucengdes.2026.114798","url":null,"abstract":"<div><div>This article is part of a series devoted to the analysis of pool thermal-hydraulic phenomena in heavy-liquid metal systems, combining experimental investigations with numerical simulations. The reference system is the Accelerator Driven System MYRRHA, currently under construction at SCK CEN (Belgium), which couples a particle accelerator with a pool-type research reactor cooled by lead-bismuth eutectic (LBE). An extensive experimental campaign has been carried out in the E-SCAPE facility, a 1/6-scaled LBE mockup of the primary vessel of MYRRHA, equipped with additional auxiliary circuits. In parallel, numerical studies based on computational fluid dynamics (CFD) are performed with the main objective of developing and validating robust CFD methodology capable of accurately simulating large-scale heavy-liquid metal experimental facilities. These studies allow the investigation of thermal-hydraulic phenomena and provide detailed insight into thermal and momentum fields. Model validation is achieved through systematic comparison between simulations and measurements obtained in E-SCAPE for a range of operating scenarios, including forced and natural circulation, symmetric and asymmetric loss-of-flow events, and asymmetric decay heat removal conditions.</div><div>In this paper, the numerical methodology to simulate the E-SCAPE facility under both steady-state and transient operating conditions is established. Since the natural circulation condition and the transition from forced to natural circulation are characterized by a strong coupling between thermal and momentum fields, detailed numerical description of the E-SCAPE core simulator, of the upper plena regions and of the external circuits are important. The main reason is that the flow rate in the core region is driven by the balance between the pressure drop and the thermal buoyancy effect and the flow rates in the external circuits are based on pressure losses through the valves, filters, and pumps. The impact of geometric simplifications of core plates and heaters, relative to the real configuration, is evaluated on the flow and on the temperature fields.</div><div>Furthermore, the effects of numerical mesh refinement and thermal boundary conditions are investigated under both iso-thermal and thermal forced circulation operating conditions. Simulated temperature distributions and pressure losses in the pool are compared against local experimental measurements. Finally, practical guidelines are proposed for CFD simulation of forced circulation, natural circulation, and transient conditions, addressing geometry simplifications, definition of fluid and solid domains, characterization of external circuits, and numerical strategies for transient initiation. The initial conditions for transient simulations are also computed and successfully compared with experimental data.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"450 ","pages":"Article 114798"},"PeriodicalIF":2.1,"publicationDate":"2026-02-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146190733","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Selection of RANS turbulence model for calculating thermal hydraulics of fuel assemblies of LMC reactors at low Reynolds numbers","authors":"K. Sergeenko ,&nbsp;A. Krutikov ,&nbsp;L. Golibrodo ,&nbsp;A. Zefirova ,&nbsp;M. Shishlenin","doi":"10.1016/j.nucengdes.2025.114741","DOIUrl":"10.1016/j.nucengdes.2025.114741","url":null,"abstract":"<div><div>Liquid metal coolants are considered in various fast neutron reactors. This paper presents the results of DNS calculations of heat and mass transfer processes of liquid metal coolants in the following regions: a parallel-plate channel, a rod bundle at cross-flow, and a channel with a step. The regions are selected in such a way as to describe as comprehensively as possible the thermal hydraulics of the fuel assembly, in which the spacing of the rods is carried out by the spacer grids. Two Reynolds numbers (<em>Re</em> = 2500 and 5000) and three Prandtl numbers (<em>Pr</em> = 0.0043, 0.0154 and 0.0324) are considered. The Prandtl numbers considered correspond to the liquid metal coolants used in fast neutron reactors: sodium, lead and lead bismuth, respectively.</div><div>The results of DNS simulations were compared with the data obtained using various RANS turbulence models (k-ε realizable two layer, k-ω SST, k-ε EB and k-ε V2F). It is shown that the k-ε V2F turbulence model describes the thermal-hydraulic processes in the best way. It recommends use this model for thermal-hydraulic calculations of fuel assemblies in reactor plants with liquid metal coolants at low Reynolds numbers.</div><div>It is shown that for all considered RANS turbulence models, the error in determining the heat transfer increases with increasing Peclet number in the parallel-plate channel. This is due to the features of the mechanisms of turbulent heat transfer for coolants at low Prandtl number.</div><div>The temperature profile is also compared with theoretical data. Satisfactory agreement is shown for all simulations in the area up to y⁺ ≈ 60. This can be used in experimental studies focused on determining the Nusselt number.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"450 ","pages":"Article 114741"},"PeriodicalIF":2.1,"publicationDate":"2026-02-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146190732","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Erratum to “An analysis of the brittleness indices of SiC layer in the TRISO fuel,” Nucl. Eng. Design 446 (2026) 114614.","authors":"Makuteswara Srinivasan","doi":"10.1016/j.nucengdes.2026.114793","DOIUrl":"10.1016/j.nucengdes.2026.114793","url":null,"abstract":"","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"449 ","pages":"Article 114793"},"PeriodicalIF":2.1,"publicationDate":"2026-02-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146189363","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Analytic hierarchy process-based integrated assessment of IAEA'S 19 infrastructure criteria for small modular reactor technology assessment","authors":"Alif Imran Mohd Shuhaimi,&nbsp;Mohd Syukri Yahya,&nbsp;Shamsul Amri Sulaiman,&nbsp;Eli Syafiqah Aziman","doi":"10.1016/j.nucengdes.2026.114797","DOIUrl":"10.1016/j.nucengdes.2026.114797","url":null,"abstract":"<div><div>Achieving Net Zero Emissions (NZE) by 2050 necessitates a strategic combination of renewable energy sources and dependable, clean baseload power. Nuclear energy is increasingly recognized as a critical element of this transition, with recent global policy endorsements calling for a tripling of nuclear capacity, underscoring its growing importance. Substantial expansion is anticipated not only in established nuclear markets but also in newcomer countries. However, initiating nuclear power programs poses significant challenges for newcomers, particularly due to massive capital investment, long-term infrastructure commitments, and the complexity of managing associated development risks. Reactor Technology Assessment (RTA) is therefore an essential process to ensure nuclear projects are delivered on time, within budget, safely, and to specifications. The International Atomic Energy Agency (IAEA) supports this process through its comprehensive framework of 19 infrastructure issues, addressing key dimensions such as national position, technical, and economic factors. This study presents a systematic, multi-criteria decision-making framework for conducting RTA, focusing on Small Modular Reactors (SMRs) as viable options for deployment in newcomer countries. The framework integrates the IAEA's infrastructure guidance with a PESTEL-based evaluation model. A case study is conducted for Malaysia to identify suitable SMRs for a Nuclear Hydrogen Hybrid System (NHHS). The assessment framework is validated using the Analytic Hierarchy Process (AHP), ensuring robustness and consistency in decision-making. The findings provide insights into aligning SMR technology selections with national requirements and risk profiles, offering a replicable methodology to support evidence-based decision-making in emerging nuclear energy programs.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"450 ","pages":"Article 114797"},"PeriodicalIF":2.1,"publicationDate":"2026-02-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146190256","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Initial study on the effects of TRISO-matrix interactions on TRISO particle performance in ceramic matrix dispersion microencapsulated fuel pellet","authors":"Junqiang Zheng ,&nbsp;Jingyu Nie ,&nbsp;Ya'nan He ,&nbsp;Yingwei Wu ,&nbsp;Jing Zhang ,&nbsp;Yuanming Li ,&nbsp;Shichao Liu ,&nbsp;G.H. Su","doi":"10.1016/j.nucengdes.2026.114811","DOIUrl":"10.1016/j.nucengdes.2026.114811","url":null,"abstract":"<div><div>This study presents a comprehensive multi-physics coupled analysis of TRISO-based Ceramic matrix Dispersion Microencapsulated (CDM) fuels, focusing on the debonding behavior at the TRISO-matrix interface under irradiation and thermal conditions, as well as the influence of TRISO-matrix interactions on TRISO particle performance within fuel pellets. A two-dimensional axisymmetric representative unit cell model was developed to investigate the effects of interfacial debonding characteristics under varying bonding parameters. Based on these findings, a three-dimensional pellet-scale multi-physics coupled model incorporating the Cohesive Zone Model was established to simulate interfacial debonding between TRISO particles and the matrix and its impact on structural integrity and failure probability. The model integrates irradiation-induced deformation, thermomechanical behavior, fission gas release, and interfacial damage evolution, enabling detailed evaluation of stress distribution and failure probability in both TRISO particles and SiC matrices.</div><div>Key results reveal that early-stage debonding at the TRISO-matrix interface is primarily driven by thermal expansion mismatch, with irradiation-induced deformation of the PyC layer contributing minimally. Smaller initial damage displacements accelerate interface failure, while larger failure displacements delay debonding but increase coating layer stresses, particularly in the SiC layer. At low temperatures, internal gas pressure within TRISO particles remains insufficient to induce critical tensile stress in the SiC layer, whereas PyC layer irradiation-induced deformation dominates the failure probability of the SiC coating. Additionally, spatial particle distribution and the surrounding matrix state significantly influence SiC coating failure, with localized high tensile stresses in narrow matrix regions posing risks for microcrack initiation.</div><div>Comparative analysis between pellet-scale CDM models and single TRISO particle models highlights the amplification of PyC layer hoop stresses due to particle-matrix interactions. The maximum failure probability of SiC coatings in pellet-scale models is approximately one order of magnitude higher than in single-particle simulations, underscoring the critical role of matrix confinement and interparticle interactions. These findings emphasize the necessity of considering high-packing-fraction configurations and interfacial bonding parameters in TRISO-based fuel performance assessments. Future work will focus on quantifying the effects of particle-matrix interactions on TRISO particle performance.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"450 ","pages":"Article 114811"},"PeriodicalIF":2.1,"publicationDate":"2026-02-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146190259","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}