Nuclear Engineering and Design最新文献

{"title":"Application of an efficient SPOD method to the flow characteristics observation of 2 × 2 fuel rod bundles","authors":"Yunhao Zhang,&nbsp;Chaojie Xing,&nbsp;Haifan Liao,&nbsp;Xinying Wang,&nbsp;Qiang Li,&nbsp;Haijun Wang","doi":"10.1016/j.nucengdes.2025.114438","DOIUrl":"10.1016/j.nucengdes.2025.114438","url":null,"abstract":"<div><div>A fast spectral proper orthogonal decomposition (F-SPOD) method is proposed for the efficient analysis of large-scale flow datasets and is applied to investigate unsteady flow structures around an exposed 2 × 2 nuclear fuel rod bundle. Spectral proper orthogonal decomposition (SPOD) is widely used to extract coherent structures across different frequencies, particularly in periodic flow phenomena such as vortex shedding and coolant-induced instabilities. However, classical SPOD becomes computationally expensive in three-dimensional cases due to the high spatial resolution required. To overcome this limitation, F-SPOD first applies singular value decomposition (SVD) to the data matrix to extract orthogonal spatial modes and corresponding temporal coefficients. Spectral analysis is then performed in the reduced temporal space, significantly improving computational efficiency and memory usage. Validation results confirm that F-SPOD retains excellent accuracy compared to classical SPOD while dramatically reducing computational cost. When applied to the fuel rod case, F-SPOD reveals that high-frequency pressure fluctuations dominate near the inlet and evolve into oblique low-frequency waves downstream. Tangential velocity fluctuations are concentrated near wall surfaces and exhibit a rotational pattern along the flow direction. Additionally, the analysis highlights the importance of monitoring point density: insufficient spatial sampling can lead to oversmoothed modal energy spectra and the omission of critical frequency components, thereby reducing the reliability of modal interpretations. The adoption of the F-SPOD algorithm is thus crucial for improving computational efficiency in such analyses.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114438"},"PeriodicalIF":2.1,"publicationDate":"2025-09-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145221232","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":"Carburization behavior and tensile properties of superalloys Inconel 617 and Incoloy 800H for very-high-temperature gas-cooled reactors under graphite dust-coated environment at 950 °C","authors":"Bin Du,&nbsp;Jiahao Chang,&nbsp;Penghui Xiao,&nbsp;Lei Peng,&nbsp;Huaqiang Yin,&nbsp;Xuedong He,&nbsp;Tao Ma","doi":"10.1016/j.nucengdes.2025.114489","DOIUrl":"10.1016/j.nucengdes.2025.114489","url":null,"abstract":"<div><div>The carburization mechanisms and changes in tensile properties of Inconel 617 and Incoloy 800H following a 100-hour corrosion exposure in an environment coated with graphite dust at 950 °C was studied in this paper. Through the microstructural analysis and mechanical property testing, it was discovered that Inconel 617 undergoes carbon atom inward diffusion, forming chromium/molybdenum carbide precipitates, which result in a reduction in ultimate tensile strength of 25.6 % and a decrease in total elongation of 64.3 %, leading to brittle fracture. In contrast, Incoloy 800H exhibits minimal degradation in mechanical properties due to the surface Si element suppressing carbon diffusion. Further research demonstrates that pre-oxidation treatment can form a dense Cr₂O₃ layer on the surface of Inconel 617, reducing its carburization by 63.4 % and restoring its mechanical properties to levels close to the original. These findings provide critical guidance for material selection and corrosion prevention strategies in VHTR applications.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114489"},"PeriodicalIF":2.1,"publicationDate":"2025-09-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145221205","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 comparative study of oxide scales formed on Alloy 690 in deaerated supercritical water and supercritical CO2 at 600 °C","authors":"Jiamei Wang,&nbsp;Chen Peng,&nbsp;Kai Chen,&nbsp;Xianglong Guo,&nbsp;Lefu Zhang","doi":"10.1016/j.nucengdes.2025.114502","DOIUrl":"10.1016/j.nucengdes.2025.114502","url":null,"abstract":"<div><div>Selecting an optimal material that offers a balanced combination of mechanical strength, outstanding corrosion resistance, and minimal neutron absorption remains a key challenge for both Generation-IV nuclear systems and materials science research. Austenitic alloys, including Fe-based stainless steels and Ni-based alloys, have emerged as promising alternatives to ferritic/martensitic (F/M) steels, owing to their superior corrosion resistance and improved creep performance. Among them, high-chromium Ni-based alloys demonstrate superior corrosion and oxidation resistance in high temperature steam-exceeding that of Fe-based austenitic stainless steels and F/M steels by over an order of magnitude. In this work, a comparative study of oxide scales formed on Alloy 690 in deaerated supercritical water and supercritical CO₂ at 600 °C was conducted. The study found that weight gains in both environments follow near-cubic rate laws. The superior oxidation resistance observed in both environments, compared to other pure austenitic alloys such as 800H, 316L, and 347, can mainly be attributed to its high chromium content. Nearly 1.5 times higher oxidation rate was observed in supercritical CO₂ than in SCW. A key observation was that the direct external oxidation and the rapid transition from internal to external oxidation within the initial 24-hours exposure in both two environments are responsible for its superior oxidation resistance. With prolonged exposure time, Cr-rich spinel oxides and Ni-rich networks within the internal oxidation zone gradually convert into Cr₂O₃, contributing to the growth of this protective chromia scale and thereby significantly retarding the oxidation process. Severe grain boundaries (GB) migration and oxidation was observed in both environments as the Cr-rich oxides at the GB was less effective at preventing oxidation, resulting in a significantly faster oxidation rate in these areas compared to the bulk grains. The slightly higher oxidation rate in supercritical CO₂ might be mainly attributed to the limited protection provided by the sparse outer NiO oxide scale and the breakdown of the innermost SiO₂ rich oxide film at O/M interface.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114502"},"PeriodicalIF":2.1,"publicationDate":"2025-09-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145221208","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":"Autonomous decision-making of operational schedules for lead–bismuth fast reactors based on BOHB optimization","authors":"Ke Su ,&nbsp;Shouyu Cheng ,&nbsp;Genglei Xia ,&nbsp;Haochen Ma","doi":"10.1016/j.nucengdes.2025.114476","DOIUrl":"10.1016/j.nucengdes.2025.114476","url":null,"abstract":"<div><div>This study proposes a hybrid autonomous decision-making framework for Lead-Bismuth Eutectic (LBE) Cooled Fast Reactors (LBEFRs), designed for deployment in remote and maritime environments where real-time human intervention is limited. To address challenges posed by uncertain missions, variable loads, and fault scenarios, the framework integrates Bayesian Optimization with Hyperband (BOHB) to jointly optimize discrete operational schedules and continuous control targets in a high-dimensional, mixed-integer space. A backpropagation neural network (BPNN) surrogate model is employed to approximate thermal–hydraulic behavior with minimal computational overhead, enabling real-time decision-making. The framework targets fault-tolerant conditions in which the reactor retains partial operability, aiming to maintain system functionality rather than replicate conventional safety responses. It is designed to complement, rather than replace, traditional safety systems in mission-critical scenarios. The framework’s performance is validated under two representative fault conditions: a steam line rupture and a turbine overspeed event, where it autonomously derives reconfiguration strategies that stabilize system parameters while satisfying safety constraints. Results demonstrate strong convergence, high accuracy, and robust adaptability, confirming the framework’s effectiveness for operational strategy optimization in LBEFRs and its potential for application in next-generation intelligent nuclear systems.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114476"},"PeriodicalIF":2.1,"publicationDate":"2025-09-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145221206","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 modelling and preliminary analysis on the freeze valve behaviour in a Molten Salt Fast Reactor","authors":"S. Deanesi ,&nbsp;A. Cammi ,&nbsp;S. Lorenzi","doi":"10.1016/j.nucengdes.2025.114486","DOIUrl":"10.1016/j.nucengdes.2025.114486","url":null,"abstract":"<div><div>In the Molten Salt Fast Reactor (MSFR), the safety function related to criticality and cooling is also delivered via draining tanks developed to accommodate liquid fuel discharged by gravity during normal maintenance or accidental events. In particular, a safety barrier that is under investigation to manage accidental scenarios consists of a salt-frozen plug designed to melt in case of loss of power or overheating. Specifically, the salt plug should melt as a consequence of an unintended temperature increase, allowing the salt to drain into a safety tank. The mechanism based on the state of the plug embraces the passive safety concept, highlighting the importance of delving into the steady state of the freeze valve and the transient behaviour. This paper proposes a preliminary analysis of the behaviour of the freeze valve within a domain that represents a symmetric portion of the MSFR primary loop. The model, developed in OpenFOAM, couples melting and solidification phenomena to a multiphysics solver, coupling neutronics and thermal-hydraulics. The 3D domain represents 1/16th of the MSFR fuel loop and is equipped with a cylindrical region that mimics the presence of the freeze valve. Two scenarios are considered to assess the impact on the freeze plug behaviour of initial and boundary conditions meant to represent different cooling strategies. When the plug is coupled with the MSFR fuel loop, the selection of boundary and initial conditions strongly affects the plug melting time. This highlights the need for design specifications for the geometry and the cooling mechanisms to correctly operate the freeze valves.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114486"},"PeriodicalIF":2.1,"publicationDate":"2025-09-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145221204","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":"Performance characterization of an axial closed Brayton cycle compressor operating with helium-nitrogen gas mixture","authors":"Arif Aziz ,&nbsp;Qun Zheng ,&nbsp;Adil Malik ,&nbsp;Ghulam Ishaque ,&nbsp;Muhammad Haris ,&nbsp;Phengphanh Taikeophithoun","doi":"10.1016/j.nucengdes.2025.114496","DOIUrl":"10.1016/j.nucengdes.2025.114496","url":null,"abstract":"<div><div>This study evaluates the possible use of a mixture of helium-nitrogen as a working medium in axial compressors employed in terrestrial nuclear facilities. As helium in its pure form is considered to be an excellent coolant owing to its superior transport capabilities, nonetheless it poses challenges in compression. With this, the high temperature gas cooled reactor (HTGR) energy systems result in increased size, greater mass, elevated costs, and dynamic issues in turbomachinery. This research provides the detailed investigation of thermophysical and transport properties of helium and nitrogen binary gas mixture at specified conditions which demonstrate that the heat transfer coefficient enhances by 4.2 % when its molar weight equals 13 g/mole. Therefore, a new helium-nitrogen compressor is subsequently designed. The performance evaluation is done, and the results indicate that just 22.86 % of the stages in the helium compressor are necessary to attain the requisite pressure in a helium-nitrogen compressor. The HTGR operating on the closed Brayton cycle (CBC) has decreased its compressor stages from 16 to 4 only. As a result, helium-nitrogen binary gas mixture is more appropriate for use in the turbocompressors of HTGR power plants than pure helium.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114496"},"PeriodicalIF":2.1,"publicationDate":"2025-09-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145155936","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":"Static-dynamic coupled analysis of pumping station structure of a nuclear power plant using viscoelastic static-dynamic unified artificial boundary","authors":"Yin Xunqiang ,&nbsp;Zhao Min ,&nbsp;Yang Weilong ,&nbsp;Zhang Junkai ,&nbsp;Li Jianbo","doi":"10.1016/j.nucengdes.2025.114417","DOIUrl":"10.1016/j.nucengdes.2025.114417","url":null,"abstract":"<div><div>The pumping station providing essential cooling water circulation are critical components in nuclear power plants (NPPs) throughout China. It is of great significance to investigate the seismic behavior of the pumping station structure considering complex heterogeneity site and load conditions according to the nuclear safety design requirements. In this study, a novel viscous-spring boundary, which is a more efficient and accurate methodology, is proposed and implemented in the ANSYS finite element software to study the seismic safety of the pumping station structure of NPPs. The Viscoelastic Static-Dynamic Unified (VSDU) artificial boundary based on Newmark’s integration scheme is proposed for nonlinear analysis of structures under coupled static-dynamic excitation. The User Programmable Features (UPFs) available in ANSYS were employed to integrate the proposed analysis method. Also, a complete toolkit for general static-dynamic coupled analysis is created using scripting-based ANSYS Parametric Design Language (APDL) and Graphical User Interface (GUI). Three validation examples are presented to demonstrate the accuracy and efficiency of the proposed method. Finally, seismic safety analysis of pumping station structure of NPPs based on the actual complex site conditions in China is also presented to illustrate the analysis procedure, and the results demonstrate that the proposed technique of investigating the seismic responses and stability of pumping station structure can provide more objective indices to evaluate the seismic safety during safety shutdown earthquake.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114417"},"PeriodicalIF":2.1,"publicationDate":"2025-09-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145155929","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":"Heat pipe failure accident analysis of the multipurpose dual-mode heat pipe nuclear reactor power system","authors":"Panxiao Li,&nbsp;Haocheng Zhang,&nbsp;Zhipeng Zhang,&nbsp;Chenglong Wang,&nbsp;Kailun Guo,&nbsp;Wenxi Tian,&nbsp;Suizheng Qiu","doi":"10.1016/j.nucengdes.2025.114471","DOIUrl":"10.1016/j.nucengdes.2025.114471","url":null,"abstract":"<div><div>This study conducts a comprehensive analysis of heat pipe failure accidents in the Multipurpose Dual-Mode Heat Pipe Nuclear Reactor power system to validate its thermal safety characteristics. Considering the importance of heat pipe reliability in reactor operations, the research focuses on three primary failure modes—heat transfer failure, condensation failure, and detachment failure—under scenarios where up to 5 % of heat pipes fail simultaneously. Utilizing a hybrid methodology integrating nuclear-thermal coupling simulations via OpenMC and finite element analysis (FEA) in COMSOL, the power distribution of fuel assemblies and full-power operational characteristics were quantified. A three-dimensional thermal–hydraulic model incorporating thermal resistance networks was developed to simulate heat pipe performance. Results demonstrate that under reference operating conditions, all core components, including fuel rods (peak temperature 1038.6 K), cladding (1017.5 K), and heat pipe walls (979.2 K), remain significantly below safety thresholds. In the event of simultaneous failure of the three highest-power heat pipes (6.67 % of total), the most severe temperature increments (up to 26.65 % rise in peak fuel temperature) occur during detachment failure, yet all components sustain temperatures within prescribed limits. Notably, cascading failures are mitigated as residual heat pipes maintain operational integrity, with maximum heat transfer capacities remaining below design limits (8241.7 W). The study confirms the robustness of the reactor’s thermal redundancy and heat pipe layout, ensuring safe operation even under worst-case failure scenarios. These findings provide critical insights for the advancement of heat pipe-cooled nuclear reactor (HPR) designs.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114471"},"PeriodicalIF":2.1,"publicationDate":"2025-09-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145155938","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":"Optimization of the primary frequency regulation ability of large-scale pressurized water reactor","authors":"Muping Li,&nbsp;Haomin Wang,&nbsp;Peiwei Sun,&nbsp;Xinyu Wei","doi":"10.1016/j.nucengdes.2025.114498","DOIUrl":"10.1016/j.nucengdes.2025.114498","url":null,"abstract":"<div><div>The frequency regulation requirement rises with an increasing proportion of clean energy in power grid operation. Power grid operators hope nuclear power plants will participate in Primary Frequency Regulation (PFR) to reduce the burden of traditional frequency-regulation power plants. A simulation platform for a typical Pressurized Water Reactor (PWR) and power grid was developed using MATLAB/Simulink to analyze the PFR characteristics of nuclear power plants. An optimal dead zone of ±0.025–0.05 Hz and a differential adjustment coefficient of 2.5–3 % were determined, while the integration of a 0–30 % maximum frequency modulation step in the PFR function and a feedforward coefficient below 300 in main steam valve control was found to significantly enhance frequency regulation. These results provide a reference for optimizing PFR strategies in PWR nuclear power plants.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114498"},"PeriodicalIF":2.1,"publicationDate":"2025-09-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145155930","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 control strategy for supercritical carbon dioxide direct cooled nuclear power system","authors":"Yunzhi Chai ,&nbsp;Jiashuang Wan ,&nbsp;Kai Xiao ,&nbsp;Zhengxi He ,&nbsp;Zhi Chen ,&nbsp;Shifa Wu","doi":"10.1016/j.nucengdes.2025.114493","DOIUrl":"10.1016/j.nucengdes.2025.114493","url":null,"abstract":"<div><div>The supercritical carbon dioxide (S-CO₂) direct cooled nuclear power system is an energy conversion system that utilizes S-CO₂ as the medium to transform nuclear thermal energy into electrical or mechanical power through a direct cycle. It offers advantages such as system simplification, compact structure, high maneuverability, and superior thermal efficiency. However, the system poses significant control challenges due to its characteristics of strong multi-equipment coupling, wide parameter variations, and complex physical processes. To address the coordination of highly variable external load demands, it is important to investigate its control strategy. In this paper, a simulation model of the S-CO₂ direct cooled nuclear power system is established on the MATLAB/Simulink. Through dynamic characteristic analysis under typical operational modes, including step transient conditions, linear transient conditions, and load rejection transient conditions, a hybrid throttling-bypass control strategy is proposed. Specifically, the throttling control subsystem is designed to manage step transient conditions and linear transient conditions requirements, while the bypass control subsystem is designed for load rejection transient conditions. Simulation results demonstrate that the proposed control strategy achieves good control performance for the S-CO₂ direct cooled nuclear power system. This paper provides valuable reference for the design of control systems in S-CO₂ direct cooled nuclear power system.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114493"},"PeriodicalIF":2.1,"publicationDate":"2025-09-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145155937","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}