Annals of Nuclear Energy最新文献

{"title":"Optimization of core design to achieve ultra-long core life in alternative ways for modular gas cooled fast reactor","authors":"Shohanul Islam","doi":"10.1016/j.anucene.2025.111241","DOIUrl":"10.1016/j.anucene.2025.111241","url":null,"abstract":"<div><div>This study explores three alternative approaches to identify the suitable core design for achieving ultra-long core life in the Allegro-50 MW_th small modular gas-cooled fast reactor. The approaches considered include: the strategic use of radial blanket, the incorporation of the heterogeneous distribution of fertile and fissile materials alongside a radial blanket, and the application of a fertile material coating around the fuel region of pin cells combined with heterogeneous pin cells. Neutronics analysis was carried out using the OpenMC Monte Carlo code, and a comprehensive depletion analysis revealed that all proposed models maintained criticality for extended periods. Both the Radial model (Model 2 (R)) and the Pin Heterogeneous model (Model 3 (Pin_H)) achieved a stable effective multiplication factor after an initial decline, maintaining this stability for a century. The mass evolution of these two models also showed an increase in the inventory of the key fissile isotope (Pu-239) over time. However, this stability and ultra-long core life came at the cost of non-uniform neutron flux and power distribution at BOL, which impacted the power peaking factor, especially for Model 2 (R) and Model 3 (Pin_H). Despite this, all models maintained the power peaking factor within acceptable limits and exhibited a more uniform distribution of neutron flux and power distribution at EOL. Although the Coated Fertile Mixture model (Model 4 (Mix)) did not achieve a stable multiplication factor like the other two proposed models, it still attained a longer core life as the effective multiplication factor decreased more gradually than the reference model (Model 1 (Ref)). Additionally, Model 4 (Mix) achieved a more uniform distribution of neutron flux and power distribution. All proposed models demonstrated satisfactory beta effective values, negative Doppler constants, control rod worth, and positive shutdown margin with similar characteristics in terms of neutron energy spectra. While all three models yielded successful results, Model 3 (Pin_H) emerged as the most favorable option due to its satisfactory performance across all neutronics parameters.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"215 ","pages":"Article 111241"},"PeriodicalIF":1.9,"publicationDate":"2025-02-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143394463","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":"Core-wide multi-physics simulation of fuel behaviour during large-break LOCA using NEXUS","authors":"Aiden Peakman ,&nbsp;Robert Gregg ,&nbsp;Victor Martinez-Quiroga ,&nbsp;Jordi Freixa","doi":"10.1016/j.anucene.2025.111228","DOIUrl":"10.1016/j.anucene.2025.111228","url":null,"abstract":"<div><div>Large-break loss-of-coolant accidents (LB-LOCAs) are among the most critical scenarios in nuclear power plant safety analysis due to their potential to significantly challenge the integrity of barriers confining radioactive material. This study demonstrates, for the first time, the application of the NEXUS framework to perform detailed, core-wide fuel assessments during a design basis accident. NEXUS integrates RELAP for thermal-hydraulics, PARCS for neutronics and ENIGMA for fuel performance to analyse the thermo-mechanical responses of fuel rods across the reactor core during accident conditions. The analysis, applied to a 1240 MWe/3400 MWth Pressurised Water Reactor (PWR) with Optimized ZIRLO cladding and conventional UO<span><math><msub><mrow></mrow><mrow><mn>2</mn></mrow></msub></math></span> fuel, presents a computationally efficient methodology to capture core-wide fuel behaviour — an area where comprehensive approaches remain scarce in the literature, particularly for metrics beyond peak cladding temperature.</div><div>An LB-LOCA scenario, specifically a double-ended guillotine break of a cold leg, was modelled as a design basis accident to evaluate fuel integrity under limiting conditions. Four LB-LOCA cases were simulated to assess the impact of varying power profiles and fuel states on cladding rupture stress and plastic hoop strain increment. Enhanced features in the ENIGMA code, including high-temperature clad creep, oxidation, and failure models, as well as a dynamic phase change model, were employed to facilitate margin-to-failure calculations under LOCA conditions. Results highlight relatively benign outcomes at the beginning of cycle (BOC) compared to more challenging end-of-cycle (EOC) scenarios, primarily driven by variations in rod internal pressures.</div><div>Both best-estimate and power conservatism cases were studied to assess fuel performance. The successful demonstration of NEXUS’ capabilities opens avenues for employing best-estimate plus uncertainty (BEPU) methodologies in future evaluations. Given its computational efficiency, this study highlights the NEXUS framework’s ability to balance detailed fuel assessments with reasonable computational demands for design basis accidents.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"215 ","pages":"Article 111228"},"PeriodicalIF":1.9,"publicationDate":"2025-02-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143387332","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"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":"Utilization of NEUP Data for HTGR Thermal-Fluid Code Validation: A New Resource Platform","authors":"Sunming Qin,&nbsp;Gerhard Strydom","doi":"10.1016/j.anucene.2025.111219","DOIUrl":"10.1016/j.anucene.2025.111219","url":null,"abstract":"<div><div>The high-temperature gas-cooled reactor (HTGR) is a Generation-IV advanced nuclear reactor design that has received significant attention due to its ability to generate high-temperature heat ranging from 750-950 °C. HTGR designs feature tri-structural isotropic (TRISO) fuel coated with ceramics, including carbon and silicon carbide, which allows for high-temperature operation while maintaining structural and fission product barrier integrity. Since its inception in 2009, the U.S. Department of Energy (DOE) Office of Nuclear Energy’s Nuclear Energy University Program (NEUP) invested more than $500 million in U.S. university nuclear research, specifically concentrating on advancing advanced reactor technologies, materials, and fuel cycles. By fiscal year 2024, the NEUP funded 36 projects dedicated to HTGR thermal-fluid (TF)<span><span>¹</span></span> research at 13 universities, each contributing significantly to our understanding of this technology. The outcomes of these diverse projects have been disseminated through dissertations and theses, final NEUP reports, peer-reviewed journal articles, and presentations at academic conferences, forming a comprehensive tapestry of knowledge. Despite the substantial value of these projects, their public domain dissemination has been fragmented, posing challenges for accessibility to researchers and policymakers and leading to underutilization of DOE investments. Recognizing this critical gap and its potential value for the future of nuclear research, the DOE’s Advanced Reactor Technologies (ART) Gas-Cooled Reactor (GCR) program has been conducting an extensive survey of completed and ongoing HTGR NEUP projects. This survey enabled the compilation of crucial data, resulting in the development of a new public-access data resource tailored for computational fluid dynamics and system code verification and validation (V&amp;V), specifically designed for HTGR TF applications.</div><div>Additionally, the data collection process during the survey process has revealed a significant challenge in central data organization, due to individual researchers from different institutes employing varying standards and preferences for recording and documenting experimental data. Therefore, an urgent need has been identified to establish a standardized reporting format for HTGR experimental projects. Addressing this issue is essential for enhancing collaboration, maximizing the impact of DOE investments, and ensuring the advancement of HTGR research and development (R&amp;D).</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"215 ","pages":"Article 111219"},"PeriodicalIF":1.9,"publicationDate":"2025-02-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143387231","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"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":"High-burnup boiling water reactor steady-state operating conditions and fuel performance analysis","authors":"Nathan Capps ,&nbsp;Robert Salko ,&nbsp;Mehdi Asgari ,&nbsp;Aaron Wysocki ,&nbsp;Ian Greenquist ,&nbsp;Shane Henderson ,&nbsp;Baris Sarikaya ,&nbsp;James Tusar ,&nbsp;Ian Porter","doi":"10.1016/j.anucene.2025.111247","DOIUrl":"10.1016/j.anucene.2025.111247","url":null,"abstract":"<div><div>The primary operational costs for existing nuclear reactors are plant operation costs, maintenance costs, and fuel costs, all of which are influenced by the materials used and the design of the reactor core. Optimizing core design parameters—including burnup limits and enrichment levels—can lengthen cycles, reduce outages, reduce reload batch fractions and spent fuel storage requirements, and lower maintenance and operating expenses, thereby enhancing economic viability. Furthermore, developing higher-fidelity tools to simulate these parameters enables better identification of the available margin, improves overall plant safety, and improves the understanding a given plant’s responses to accident scenarios. In the US, much of the research and development focus has traditionally been on pressurized water reactors (PWRs), but boiling water reactors (BWRs) comprise approximately one-third of the US reactor fleet. Modeling and simulation advances for BWRs and PWRs—particularly those achieved through the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program—are crucial to the long-term viability of the light–water reactor industry. A key research area of the high burnup and increased enriched fuel initiative is focused on addressing issues related to postulated loss-of-coolant accident (LOCA) scenarios. NEAMS has dedicated significant effort to enhancing tools to better support BWRs. A current focus is showcasing the BWR framework for high-burnup LOCA analysis. This high-fidelity steady-state analysis is a first step toward demonstrating a best-estimate, pin-by-pin high-burnup BWR LOCA analysis to assess full-core cladding rupture behavior for a representative BWR. The objective of this effort is to provide a modeling capability that will help elucidate and provide a best-estimate evaluation for cladding rupture susceptibility in BWRs. This modeling capability could then be used to prevent and/or mitigate cladding ruptures in postulated accident scenarios without penalizing operational parameters. Additionally, the results of this work will help identify strategies for finding additional margins or potentially limiting cladding ruptures through core design optimizations to enable more efficient core designs.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"215 ","pages":"Article 111247"},"PeriodicalIF":1.9,"publicationDate":"2025-02-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143387232","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":"Optical and spectral features of new Borax/ Li2O glass systems and calculation of radiation shielding attenuation parameters","authors":"A.M. Abdelmonem ,&nbsp;Nesreen R. Abd Elwahab ,&nbsp;E.M. Abou Hussein","doi":"10.1016/j.anucene.2025.111232","DOIUrl":"10.1016/j.anucene.2025.111232","url":null,"abstract":"<div><div>Borax/Li₂O glass compositions were prepared by the common melting-quenching technique. The XRD pattern displayed the amorphous natures and the glassy states of the samples. The detected UV absorption peaks at 225 and 256 nm revealed a slight increase in absorbance intensity after exposing glasses to 60 kGy of gamma radiation, in addition to shifting of the cut-off peak from 270 to 282 nm. Optical energy band gap (Eopt) and density values exhibit a similar behavior either because of the continual addition of Borax content or by irradiation. Infrared transmittance spectra (FTIR) displayed the main vibrational bands of tetrahedral BO₄ groups at 800–1200 cm⁻¹ and the vibrational bands of trigonal borate units BO₃ at 1200–1600 cm⁻¹. After irradiation, the spectra revealed an acceptable structural stability against irradiation. Optical and structural studies obtain a positive effect on the desired glass features by the increase of the former boron content at the expense of the decrease of the modifier Li₂O content. Different theoretical radiation shielding parameters were investigated for Borax glass composites using Phy-X/PSD, MCNP-4C2, MRCScal, NGcal, Py-MBLUF, Auto-Z_eff, and GRASP programs, such as mass attenuation coefficient (MAC), half value layer (HVL), effective atomic number (Z_eff), effective conductivity (C_eff), the equivalent atomic number (Z_eq)_, the effective electron density (N_eff), absorption gamma dose rate (Dr), and the transmission factors for gamma-ray, fast, and thermal neutrons (TF). Fast neutron removal cross section (Σ_R), macroscopic removal cross section (MRCS), and macroscopic cross section of slow and epithermal neutrons were also calculated. The overall shielding parameters revealed agreement in results between the used programs, where increasing the concentration of Borax caused an improvement in MAC values and increasing of FNRCS, MRCS, and Σ_(slow), and Σ_(epithermal) for fast, slow, and epithermal neutrons were observed. The obtained results for optical, structural, and simulation shielding indicate that the glass containing the highest borax content has the highest optical, structural stability, Eopt, and density values, as well as the highest attenuation parameters for both gamma-rays and neutrons at different energies. Dr values increase as the Borax percentage concentration content increases. The study recommends the hopeful use of the investigated Borax/Li₂O glasses as transparent materials in radiotherapy and diagnostic applications.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"215 ","pages":"Article 111232"},"PeriodicalIF":1.9,"publicationDate":"2025-02-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143379073","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":"Advanced full-core modeling of fission product release in pebble-bed high-temperature gas-cooled reactors","authors":"Chenghao Cao,&nbsp;Junyi Chen,&nbsp;Jingang Liang,&nbsp;Chuan Li,&nbsp;Jianzhu Cao","doi":"10.1016/j.anucene.2025.111240","DOIUrl":"10.1016/j.anucene.2025.111240","url":null,"abstract":"<div><div>Accurate modeling of fission product release is essential for the safety of pebble-bed high-temperature gas-cooled reactors. Traditional models oversimplify by neglecting non-uniform temperature fields, variable nuclide production rates, and the randomness of pebble flow. This study introduces an enhanced diffusion calculation method that incorporates variable production rates from the burnup equation and non-uniform temperature profiles from steady-state heat conduction models. A rapid full-core release calculation refined to individual pebble dynamics is developed to better capture equilibrium core behavior. Comparing the refined model to traditional methods, results show stepwise accumulation of concentrations, reduced release rates from particles and pebbles, and increased concentration gradients within the graphite matrix. Release rate distributions for short-lived nuclides are influenced by both core temperature and neutron flux distributions, while those for long-lived nuclides are primarily temperature-dependent. The variable-temperature model indicates that, compared to uniform particle temperature models, a greater diffusion distance is required for nuclide release.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"215 ","pages":"Article 111240"},"PeriodicalIF":1.9,"publicationDate":"2025-02-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143379074","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":"A fuzzy logic-based coordinated adaptive control method for megawatt novel nuclear power systems","authors":"Qingfeng Jiang,&nbsp;Pengfei Wang","doi":"10.1016/j.anucene.2025.111245","DOIUrl":"10.1016/j.anucene.2025.111245","url":null,"abstract":"<div><div>The megawatt novel nuclear power system (MNNPS) employing heat-pipe reactors and sCO₂ Brayton cycles has significant applications. However, the unattended operating environment places high demands on the adaptive capability. In this paper, a fuzzy logic-based coordinated adaptive control (FL-CAC) method for MNNPS is proposed. The key to the method is to design fuzzy logic controllers with the current loop-controlled variable as the main-input and the remaining loop-controlled variables as the auxiliary-inputs, and to realize the coordinated adaptive adjustment of control parameters by comprehensively considering the operating status in control loops. The FL-CAC method was verified under typical operating conditions such as step and ramp load change transients. The simulation results demonstrates that it can significantly improve the control performance compared with the original control scheme. This study can provide a theoretical reference for the autonomous and reliable control of MNNPS which need to operate in an unattended environment.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"215 ","pages":"Article 111245"},"PeriodicalIF":1.9,"publicationDate":"2025-02-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143387333","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 practicalities of producing a nuclear weapon using high-assay low-enriched uranium","authors":"P. Cosgrove,&nbsp;N. Read","doi":"10.1016/j.anucene.2025.111235","DOIUrl":"10.1016/j.anucene.2025.111235","url":null,"abstract":"<div><div>It was recently argued by Kemp et al. that HALEU (high-assay low-enriched uranium, or uranium enriched up to 19.75%) can conceivably be used to produce a nuclear weapon and on this basis civilian enrichment limits should be lowered to 10% or 12% (Scott Kemp et al., 2024). We find their argument unconvincing in several respects.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"215 ","pages":"Article 111235"},"PeriodicalIF":1.9,"publicationDate":"2025-02-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143379075","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"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":"Comparison between EDF MAAP5.04 and ASTECV3 codes on a hypothetical severe accident on the ELSMOR project E-SMR design","authors":"Jeremy Bittan ,&nbsp;Nikolai Bakouta ,&nbsp;Julie-Anne Zambaux ,&nbsp;Laure Carénini","doi":"10.1016/j.anucene.2025.111258","DOIUrl":"10.1016/j.anucene.2025.111258","url":null,"abstract":"<div><div>This paper presents a comparison between EDF MAAP 5.04 and ASTECv3 codes for a hypothetical severe accident leading to core degradation in the ELSMOR project’s proposed E-SMR (Small Modular Reactor) design. The ELSMOR (Towards European Licensing of Small Modular Reactors) project was a Horizon 2020 Euratom project that ended in 2023. The consortium included 15 partners from 8 European countries, involving research institutes, major European nuclear companies, and technical support organizations. The 3.5-year project, launched in September 2019, investigated selected safety features of Light-Water (LW) SMRs with a focus on licensing aspects.</div><div>The Modular Accident Analysis Program (MAAP) is a deterministic code owned and licensed by the Electric Power Research Institute (EPRI) that simulates the response of light water moderated nuclear power plants during accidental transients. ASTEC v3, developed by IRSN, is a severe accident system code used to evaluate major nuclear accidents for different nuclear installations, focusing on western light water reactor designs.</div><div>This study compares EDF MAAP5.04 and ASTECv3 in terms of transient evolution from the initiating event (a Station Black Out), core degradation and hydrogen generation, corium relocation to the lower plenum, and in-vessel melt retention (IVMR). The comparison also includes physical phenomena in the containment, such as steam condensation on the walls. The potential for H₂ combustion, based on the specific assumptions of the selected transient, is evaluated through flammability diagrams, and a sensitivity analysis of N2 injection for inerting the containment is assessed.</div><div>This study shows that both EDF MAAP 5.04 and ASTECv3 codes can effectively simulate severe accident phenomena and mitigation strategies like IVMR, containment inerting, and reactor decay heat removal. The codes generally agree well, providing key parameter magnitudes. Some modeling improvements were made, and remaining discrepancies are discussed. Sensitivity calculations are presented to confirm the analysis and to highlight model uncertainties, particularly for steam condensation and hydrogen production.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"215 ","pages":"Article 111258"},"PeriodicalIF":1.9,"publicationDate":"2025-02-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143387331","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":"MURR LEU structural and thermal hydraulics analyses: Part II – Impacts of irradiation thermo-mechanical behavior on thermal hydraulics safety analyses","authors":"Dhongik S. Yoon ,&nbsp;Firat Cetinbas ,&nbsp;Guanyi Wang ,&nbsp;John A. Stillman ,&nbsp;Valerio Mascolino ,&nbsp;Maria Pinilla ,&nbsp;Earl E. Feldman ,&nbsp;Walid Mohamed ,&nbsp;Erik H. Wilson","doi":"10.1016/j.anucene.2025.111236","DOIUrl":"10.1016/j.anucene.2025.111236","url":null,"abstract":"<div><div>A series of structural analyses have been performed to support the conversion of the University of Missouri Research Reactor (MURR) from the use of highly enriched uranium (HEU; ≥20 wt% U-235) to low-enriched uranium (LEU; &lt;20 wt% U-235) fuel. The irradiation thermo-mechanical analysis evaluated the effects of fuel swelling, irradiation creep, thermal expansion, as well as thermal resistance from the oxide layer growth for the MURR LEU element in prototypic thermal and irradiation conditions as presented in Part I of this article. Overall, this irradiation thermo-mechanical analysis predicts smaller gap thickness reductions in previously limiting regions, and larger reductions in the middle of the outermost end channels where power density is not typically a maximum. Due to substantial differences between the channel gap reductions assumed for the previous safety analyses and those predicted by the irradiation thermo-mechanical analysis, a need to evaluate their impact on the thermal hydraulics safety analyses arose. This article presents the results from the steady-state safety analysis for normal operation as well as the two most limiting accident scenarios. The calculation models were revised in order to account for the spatial and temporal variation of the channel gap thicknesses. The results show that sufficient safety margins are still maintained for normal operation as well as during the postulated accident transients. This work provides a methodology of incorporating the irradiation thermo-mechanical behavior of plate-type fuel into the thermal hydraulics safety analyses.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"215 ","pages":"Article 111236"},"PeriodicalIF":1.9,"publicationDate":"2025-02-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143376550","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}