Nuclear Engineering and Design最新文献

{"title":"Research progress in high-temperature thermo-mechanical behaviors for modelling Cr-coated cladding under loss-of-coolant accident condition","authors":"Chunyu Yin ,&nbsp;Guanghui Su ,&nbsp;Libo Qian ,&nbsp;Qingwen Xiong ,&nbsp;Yu Liu ,&nbsp;Yingwei Wu ,&nbsp;Sijia Du ,&nbsp;Jing Zhang ,&nbsp;Zhong Xiao","doi":"10.1016/j.nucengdes.2025.114125","DOIUrl":"10.1016/j.nucengdes.2025.114125","url":null,"abstract":"<div><div>Chromium (Cr)-coated zirconium cladding has emerged as a leading candidate for accident tolerant fuel (ATF) cladding in near-term engineering applications. This cladding demonstrates enhanced resistance to high-temperature oxidation, superior mechanical properties at elevated temperatures, and a relatively high level of technological maturity. Its performance under loss-of-coolant accident (LOCA) conditions is critical to reactor safety, making it a key focus of the present study. The present work introduces an overview of research progress on high temperature thermo-mechanical behaviors for Cr-coated cladding and provides a set of fundamental safety analysis models tailored for LOCA scenarios. First, essential models for LOCA safety analysis of Cr-coated cladding are identified, including a high-temperature oxidation model (along with a Cr coating consumption model), a high-temperature creep model, a high-temperature burst model, and an embrittlement criterion. Second, based on the evaluation of experimental data from high-temperature oxidation studies, models for the growth of Cr₂O₃ layer and oxygen absorption are recommended to estimate the oxidation rate of Cr-coated cladding. Additionally, a model for Cr coating consumption is proposed. Subsequently, through a comprehensive review and reevaluation of high-temperature creep and burst data, corresponding models for Cr-coated cladding are developed respectively. Finally, embrittlement data for Cr-coated cladding are analyzed, and embrittlement criteria for both one-sided oxidation and two-sided oxidation conditions are proposed.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"439 ","pages":"Article 114125"},"PeriodicalIF":1.9,"publicationDate":"2025-05-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143913238","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 of interassembly bypass flow in Sodium Fast Reactors","authors":"A.J. Novak ,&nbsp;C. Bourdot Dutra ,&nbsp;D. Shaver ,&nbsp;E. Merzari","doi":"10.1016/j.nucengdes.2025.114044","DOIUrl":"10.1016/j.nucengdes.2025.114044","url":null,"abstract":"<div><div>Interassembly flow in Sodium Fast Reactors (SFRs) represents a bypass flow path exterior to the fuel assembly ducts. Heat transferred across this thin gap is an important component of core radial expansion, where the coupling between thermal-fluids, neutronics, and solid mechanics results in time-dependent duct bowing. These geometry changes can constitute a significant portion of the total reactivity response in transients, but are difficult to model in high-fidelity. Interassembly flow is also an important heat transfer mode during natural convection cooling. To improve our understanding of interassembly flow, this paper provides NekRS Reynolds Averaged Navier–Stokes (RANS) and Large Eddy Simulations (LES) of the interassembly flow in a 19-bundle fast reactor core. Time-averaged LES compares reasonably well with a <span><math><mi>k</mi></math></span>-<span><math><mi>τ</mi></math></span> RANS model, though RANS is not able to capture a crossflow which occurs at a large change in flow area between the duct–duct gaps and the open peripheral region. We predict velocity distributions and illustrate a multiscale postprocessing system that can be used to generate coarse-mesh closures for subchannel and porous media tools, and provide a dataset with average velocity for comparison with coarse-mesh tools.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"439 ","pages":"Article 114044"},"PeriodicalIF":1.9,"publicationDate":"2025-05-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143906224","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":"Experimental study on post-buckling behavior of fast reactor vessel under excessive earthquakes","authors":"Yiji Ye,&nbsp;Masakazu Ichimiya,&nbsp;Naoto Kasahara","doi":"10.1016/j.nucengdes.2025.114130","DOIUrl":"10.1016/j.nucengdes.2025.114130","url":null,"abstract":"<div><div>The Fukushima Daiichi nuclear accident has raised the nuclear industry’s interest in the countermeasures for Beyond Design Basis Events (BDBEs) such as excessive earthquake. Sable failure modes are acceptable in BDBEs with the safety goal being to prevent unstable failure modes. The Fast Reactor Vessel (FRV) is vulnerable to seismic buckling due to thin-walled structure. Under excessive earthquakes, the safety goal of FRV is to achieve a stable post-buckling state. This paper presents an experimental study on post-buckling behavior of short and medium cylinders under horizontal vibration, simulating the phenomena in pool and loop type FRV. Independent of buckling configuration, a global response stability is confirmed after buckling. This stability is achieved by the phase-shift phenomenon, where buckling initiation increases the frequency ratio and enables the displacement-controlled characteristic of the dynamic load. Such phenomenon is independent of input conditions. In addition, longer cylinders show a higher post-buckling frequency ratio with significant response reduction compared to short cylinders. Next, the post-buckling failure development process is investigated and can be summarized into three stages. The ultimate rupture boundary can be measured by the critical cumulative input energy, which shows clear dependency on the buckling configuration. A preliminary criterion against ultimate rupture and the energy-based failure mode map are proposed to assess the safety margin of FRV. It demonstrates a considerable margin from buckling initiation to ultimate rupture during an excessive earthquake. This paper largely extends the database and contributes to a more comprehensive understanding in the post-buckling domain of FRV under BDBEs.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"439 ","pages":"Article 114130"},"PeriodicalIF":1.9,"publicationDate":"2025-05-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143906316","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":"Optimized ensemble of neural networks for the prediction of critical heat flux","authors":"Ibrahim Ahmed ,&nbsp;Irene Gatti ,&nbsp;Enrico Zio","doi":"10.1016/j.nucengdes.2025.114111","DOIUrl":"10.1016/j.nucengdes.2025.114111","url":null,"abstract":"<div><div>Critical Heat Flux (CHF) is a thermal limit in boiling heat transfer, beyond which there is a substantial reduction in heat transfer efficiency. This phenomenon plays a vital role in the thermal engineering design of systems involving two-phase flow. As a result, an accurate CHF prediction is essential for both safety and performance, particularly in water-cooled nuclear reactors where thermohydraulic margins are critical. In this paper, a novel optimized ensemble of neural networks (NNs) for CHF prediction is proposed to enhance the accuracy of individual models trained separately with distinct architectures and hyperparameters settings. Two systematic procedures are presented to identify potentially optimal NN models and aggregate them into an optimal ensemble model. The proposed method is validated using experimental CHF data made available by the Working Party on Scientific Issues and Uncertainty Analysis of Reactor Systems (WPRS) Expert Group on Reactor Systems Multi-Physics (EGMUP) task force on AI and ML for Scientific Computing in Nuclear Engineering projects, promoted by the OECD/NEA. The results obtained show that the ensemble model outperforms standalone models and other state-of-the-art modelling approaches. Parametric and sensitivity analyses across various input parameters confirm the robustness of the ensemble model and its consistency with expected physical behaviors, further underlying its potential for improving CHF prediction in nuclear reactor applications.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"439 ","pages":"Article 114111"},"PeriodicalIF":1.9,"publicationDate":"2025-05-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143906225","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":"Radiation safety and fuel performance of pebble-bed modular high-temperature gas-cooled reactors","authors":"Yu Wang ,&nbsp;Jianzhu Cao ,&nbsp;Bing Xia ,&nbsp;Feng Xie ,&nbsp;Fu Li ,&nbsp;Haitao Wang ,&nbsp;Jiejuan Tong ,&nbsp;Yujie Dong ,&nbsp;Zuoyi Zhang ,&nbsp;Karl Verfondern ,&nbsp;Sudarshan K. Loyalka","doi":"10.1016/j.nucengdes.2025.114116","DOIUrl":"10.1016/j.nucengdes.2025.114116","url":null,"abstract":"<div><div>Safety of nuclear reactors is of wide concern in the world, and inherent safety of the reactors is the goal that the nuclear energy field has been pursuing over the last several decades. How to quantitatively evaluate the inherent safety, as well as the reactor radiation safety, is a long-term important scientific and technical issue. This study focused on the 10 MW high-temperature gas-cooled experimental reactor (HTR-10), the only operational pebble-bed modular HTGR for testing that can operate at full power currently, measured its primary coolant activity, a key indicator of reactor radiation safety, and assessed its fuel element performance, directly affecting its inherent safety feature. A method for evaluating tri-structural isotropic coated fuel particle (TRISO CFP) failure fraction and uranium contamination share was established. The release-to-birth (R/B) ratio for fission gas nuclides was &lt; 1 × 10⁻⁶, with a TRISO CFP failure fraction of 7.23 × 10⁻⁵ and uranium contamination share of 9.20 × 10⁻⁶. The TRISO CFP performance of HTR-10 surpassed that of previous HTGRs and irradiation tests conducted in the world, highlighting its excellent radiation safety and potential for large-scale commercial application of HTR-PM (that are related/based on HTR-10).</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"439 ","pages":"Article 114116"},"PeriodicalIF":1.9,"publicationDate":"2025-05-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143906317","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":"Gain scheduling model predictive control for a space thermionic nuclear reactor","authors":"Qian Ma ,&nbsp;Kai Wang ,&nbsp;Peiwei Sun ,&nbsp;Xinyu Wei","doi":"10.1016/j.nucengdes.2025.114124","DOIUrl":"10.1016/j.nucengdes.2025.114124","url":null,"abstract":"<div><div>Space thermionic nuclear reactor (STNR) has the characteristics of large delay and strong nonlinearity. It is difficult to obtain the satisfied performance with traditional control system. Model prediction control (MPC) is adopted in this study. The control law is designed and optimized through the prediction model and objective performance function. The problem of poor control performance of large delay system is solved and fast regulation is achieved. The model predictive controllers are designed at different power levels. Gain scheduling method is utilized to improve the tracking accuracy of the nonlinear system. This approach enables stable operation of STNRs across the full-power range. It is shown from the simulation analysis under the representative operational scenarios that the model predictive control system based on gain scheduling has smaller overshoot and settling time than the traditional control system. The fluctuations of major parameters are reduced significantly. Therefore, the proposed gain scheduling model predictive control provides a promising strategy to the control of STNR.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"439 ","pages":"Article 114124"},"PeriodicalIF":1.9,"publicationDate":"2025-05-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143906226","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":"Thermal-hydraulic analysis of a representative Westinghouse lead-cooled fast reactor fuel bundle using CFD","authors":"D. Wilson ,&nbsp;H. Iacovides ,&nbsp;E. Tatli ,&nbsp;C. Stansbury ,&nbsp;P. Ferroni","doi":"10.1016/j.nucengdes.2025.114117","DOIUrl":"10.1016/j.nucengdes.2025.114117","url":null,"abstract":"<div><div>Lead-cooled fast reactors (LFRs) are Generation IV reactor technologies that use molten lead as the primary coolant. Whilst lead offers advantages for economics, safety, and sustainability, its low Prandlt number and challenging experimental characteristics pose difficulties for thermal–hydraulic modelling and validation. To support LFR development, this study aims to advance modelling capabilities and understanding of the relevant physical phenomena through a series of Computational Fluid Dynamics (CFD) simulations of a Fuel Pin Bundle Simulator (FPBS) that shares design features with the Westinghouse LFR fuel assembly.</div><div>Three geometrical configurations of the FPBS have been modelled, using the Reynolds-averaged Navier-Stokes (RANS) approach: a bare pin bundle (without spacer grids), a T-junction upstream of the main test section, and the full-length 360° main FPBS test section including spacer grids and instrumentation wires. The sensitivity of the results to modelling choices, including turbulence models and approaches for the turbulent Prandtl number, is explored.</div><div>The original contributions of this study are in the assessment of different RANS models of the Reynolds stresses, the assessment of different values and functions of the turbulent Prandtl number for the modelling of the turbulent heat fluxes, the exploration of the entry conditions on the flow and thermal development along the fuel bundle and the determinations of the effects of the intrusive instrumentation on the measured quantities.</div><div>The bare bundle simulations showed only minor sensitivity to the turbulence model and produced friction factors in excellent agreement with existing correlations. Predictions in the upstream T-junction indicated the generation of significant swirl that enters the main test section, but the spacer grid acts as an effective flow straightener. Nusselt number predictions in the main FPBS test section showed good agreement with established correlations for liquid metal rod bundles. Instrumentation wires had only a minor effect on the temperature field and increased the pressure drop by 2.7 %. A sensitivity analysis of the turbulent Prandtl number (<span><math><msub><mrow><mi>Pr</mi></mrow><mi>t</mi></msub></math></span>) showed that Kay’s correlation produced Nusselt numbers that were closest to the empirical correlation of <span><span>Ushakov et al. (1977)</span></span>, with a mean deviation of 1.1 %. In contrast, a constant <span><math><mrow><msub><mrow><mi>Pr</mi></mrow><mi>t</mi></msub><mo>=</mo><mn>0.9</mn></mrow></math></span> resulted in an overprediction of 19 %.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"439 ","pages":"Article 114117"},"PeriodicalIF":1.9,"publicationDate":"2025-05-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143904594","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":"Radiation and ergonomic-aware A-star algorithm for nuclear decommissioning path planning","authors":"Justina Onyinyechukwu Adibeli ,&nbsp;Yong-Kuo Liu","doi":"10.1016/j.nucengdes.2025.114123","DOIUrl":"10.1016/j.nucengdes.2025.114123","url":null,"abstract":"<div><div>This study presents a novel modification of the A* path planning algorithm that integrates ergonomic and radiological constraints to enhance worker safety in nuclear decommissioning environments. The algorithm optimizes paths by incorporating ergonomic scores derived from biomechanical data alongside radiation exposure estimates from simulations. To assess the impact of load carriage, worker performance is analyzed under normal and weighted walking conditions, with a focus on joint biomechanics, including moment, power, angle, and velocity. Results indicate that while cumulative radiation exposure remains relatively stable, weighted walking significantly increases biomechanical strain, elevating the risk of musculoskeletal disorders (MSDs). The proposed dual-cost optimization approach provides a comprehensive framework for balancing radiation exposure and ergonomic stress, offering practical improvements for worker safety and efficiency in hazardous environments.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"439 ","pages":"Article 114123"},"PeriodicalIF":1.9,"publicationDate":"2025-05-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143904590","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":"Multiphysics modeling approach for the analysis of noble metals deposition in the Molten Salt Fast Reactor","authors":"Nicolò Iaselli ,&nbsp;Antonio Cammi ,&nbsp;Stefano Lorenzi","doi":"10.1016/j.nucengdes.2025.114054","DOIUrl":"10.1016/j.nucengdes.2025.114054","url":null,"abstract":"<div><div>Noble metals exhibit very low solubility in fluorine salts, leading to accumulation on reactor surfaces, which negatively impacts performance and safety. In this work, a new modeling capability of the OpenFOAM multiphysics solver, developed at Politecnico di Milano is proposed to analyze the deposition of noble metal fission products in the Molten Salt Fast Reactor (MSFR). To model the particle migration towards reactor walls, a tailored particle transport model and custom boundary condition were implemented. Verification against an analytical solution confirmed accuracy, followed by a sensitivity analysis on mesh refinement, which demonstrated strong dependence on wall-adjacent cell size. Simulating the reactor in full geometry and accounting for all nuclides in the salt demands high-performance computational resources, even for steady state conditions. To reduce computational effort, the deposition velocity (or mass transfer coefficient) obtained from a highly refined mesh was applied to coarser meshes using the tailored boundary conditions. This approach, combined with a single pseudo-nuclide representing the noble metals family, significantly reduces computational demand. Different mesh types were tested for steady-state reactor core simulations, showing that the deposition velocity-based strategy provides satisfactory results for the quantities of interest. Preliminary results are also presented for decay heat generated by radioactive particle deposits. The developed capability to describe noble metal behavior advances the multiphysics solver and contributes to the MSFR’s design optimization.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"439 ","pages":"Article 114054"},"PeriodicalIF":1.9,"publicationDate":"2025-05-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143902383","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":"Numerical analysis and model prediction of flow enhanced heat transfer mechanism in spirally corrugated tubes","authors":"Chuang Pan ,&nbsp;Shuhong Li ,&nbsp;Yanjun Li ,&nbsp;Jun Wu ,&nbsp;Gui Li","doi":"10.1016/j.nucengdes.2025.114113","DOIUrl":"10.1016/j.nucengdes.2025.114113","url":null,"abstract":"<div><div>The spirally corrugated tube (SCT) has advantages such as bidirectional enhanced heat transfer, which can effectively improve the economic efficiency of heat exchange equipment. However, the influence law of the start value (n) on the flow and heat transfer of the SCT is not clear enough, and it is still necessary to explore how to design and select suitable SCT according to the application working conditions. Therefore, in this study, a numerical simulation of the flow and heat transfer inside the tube of a multi-start SCT with an equivalent inner diameter (D_i = 20 mm) was carried out. The effects of n (n = 1–8), pitch ratio (p/D_i = 1.5–3.0), corrugated depth ratio (e/D_i = 0.05–0.20) and Reynolds number (Re = 5000–30000) on the velocity and temperature distributions on the multi-start SCT were investigated, and comparisons were made with four other types of enhanced tubes (conically corrugated tubes, arc-corrugated tubes, converging–diverging tubes and spirally grooved tubes). The comprehensive performance of the multi-start SCT was evaluated according to the Performance Evaluation Criteria (PEC), and the mechanism of heat transfer enhancement was revealed through the field synergy theory. The results show that the PEC of the SCT is significantly better than that of the other four types of enhanced tubes. As n and e/D_i increase, the PEC first decreases and then increases. as p/D_i increases, the PEC gradually increases. The synergy between the temperature gradient, pressure gradient and velocity of the eight-start SCT is the least affected by Re, and it can maintain a relatively high PEC, with its optimal PEC being 1.764. In addition, a prediction model for the SCT was proposed through linear fitting, and the error between the prediction model and the simulated values is within 15 %, providing guidance for the practical engineering application of the SCT.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"439 ","pages":"Article 114113"},"PeriodicalIF":1.9,"publicationDate":"2025-05-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143898949","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}