Progress in Nuclear Energy最新文献

{"title":"Exploration of grid types for radiation field representing and path-finding","authors":"Jiamei Tang , Xiaodan Li , Mengkun Li , Li Liu , Mengxiao Wang","doi":"10.1016/j.pnucene.2026.106251","DOIUrl":"10.1016/j.pnucene.2026.106251","url":null,"abstract":"","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106251"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145977699","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":"Dynamic analysis of China's load capacity factor: The roles of energy heterogeneity and economic globalization through ARDL and machine learning methods","authors":"Zhipeng Yang ,&nbsp;Yingjie Wang ,&nbsp;Jiacheng Yu ,&nbsp;Hanrui Qiu ,&nbsp;Mingjun Wang ,&nbsp;Wenxi Tian ,&nbsp;G.H. Su","doi":"10.1016/j.pnucene.2025.106222","DOIUrl":"10.1016/j.pnucene.2025.106222","url":null,"abstract":"<div><div>This study investigates the dynamic impacts of energy heterogeneity, economic globalization, and gross domestic product on China's Load Capacity Factor (LCF) within the context of its dual carbon goals. Using annual data from 1993 to 2021, the study integrates econometric and machine learning approaches to examine dynamic relationships and forecast future energy trends. The Autoregressive Distributed Lag (ARDL) model is used for empirical analysis, while the Gated Recurrent Unit (GRU) model supports forecasting. The robustness of the ARDL results is verified through cointegration tests, stability diagnostics, and Fully Modified Least Squares (FMOLS) estimation. The long-run results indicate that fossil fuel consumption has the strongest negative impact on LCF, with a coefficient of −1.110, while renewable energy consumption also negatively affects LCF with a coefficient of −0.137. In contrast, nuclear energy consumption, economic globalization, and gross domestic product contribute positively, with coefficients of +0.077, +0.564, and +0.322, respectively. In the short run, nuclear energy consumption (+0.116), renewable energy consumption (−0.275), and gross domestic product (+0.524) exhibit significant effects, whereas fossil fuel consumption and economic globalization primarily influence long-term dynamics. GRU forecasts project a substantial transformation of China's energy structure by 2030: the fossil fuel consumption share is expected to fall to 54.95 %, while nuclear and renewable energy shares are projected to rise to 7.96 % and 37.09 %, respectively. LCF is projected to stabilize overall, with a potential turning point in 2027, followed by a gradual upward trend. These findings underscore the imperative of transitioning toward a lower-carbon energy system to enhance LCF and meet sustainability goals.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106222"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145927137","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":"Study on the melting and relocation behavior of dispersion helical cruciform fuel under accidents","authors":"Shihao Wu,&nbsp;Xingyu Wang,&nbsp;Yapei Zhang,&nbsp;Yicong Lan,&nbsp;Wenxi Tian,&nbsp;Suizheng Qiu,&nbsp;G.H. Su","doi":"10.1016/j.pnucene.2025.106228","DOIUrl":"10.1016/j.pnucene.2025.106228","url":null,"abstract":"<div><div>The Helical Cruciform Fuel (HCF) has a large heat transfer area and is self-supporting, showing great potential for improving reactor efficiency and safety. However, there is still a lack of research on the melting and relocation behavior of the dispersion HCF under accident conditions. Based on the Incompressible Smoothed Particle Hydrodynamics (ISPH) method, this paper uses the Discrete Phase Model (DPM) and the Discrete Element Model (DEM) to solve the coupled motion of dispersion fuel particles and molten matrix, and further analyzes the dispersion fuel settling behavior in the melted matrix inside the cladding and the relocation behavior of matrix and dispersion fuel outside the cladding. The analysis results indicate that the settling behavior of fuel particles in the fin zone is influenced by the twisted structure, leading to local overheating of the cladding. After the failure of the cladding, the relocation states of the molten material in the cladding groove include the transition from droplet slip to stream flow, followed by the separation of small droplets, ultimately forming a narrow-wide-narrow blockage structure. This paper provides a theoretical basis and guidance for the safety assessment and accident mitigation strategy formulation of the dispersion HCF.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106228"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145927081","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":"Development of a transient numerical model for contact melting of molten material on a narrow plate component with varying contact thermal resistance","authors":"Junjie Ma,&nbsp;Wenzhen Chen","doi":"10.1016/j.pnucene.2025.106227","DOIUrl":"10.1016/j.pnucene.2025.106227","url":null,"abstract":"<div><div>During a core meltdown, the contact melting of the crust of a molten pool against narrow plate components within the reactor exhibits characteristics of a complex convective boundary condition. This complexity arises because the early stage is governed by the coupling of contact thermal resistance and heat conduction mechanisms, while the melting phase is influenced by the overall system heat transfer behavior. Thus, accurately modeling this process poses significant challenges. To address this issue, this paper establishes a transient numerical model to simulate the contact melting behavior between a heat source and a narrow plate of phase change material (PCM), based on contact melting theory. The governing equations are coupled with the heat conduction differential equations of both the heat source and the PCM, with full consideration of the role of contact thermal resistance during the conduction stage. This coupled approach avoids traditional strong numerical coupling computations, significantly reducing computational costs. The results demonstrate that the model effectively tracks the dynamic evolution of the melting rate. Furthermore, the study reveals that modifying the contact thermal resistance by introducing a eutectic reaction substantially improves the prediction accuracy of the melting initiation time, highlighting the importance of considering the influence of the eutectic reaction on contact thermal resistance in the model, which plays a key role in enhancing model accuracy.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106227"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145927082","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":"Cybersecurity risk assessment of nuclear SVDU based on attack tree","authors":"Yanqun Wu ,&nbsp;Junfeng Wang ,&nbsp;Quan Ma ,&nbsp;Mingxing Liu ,&nbsp;Meiyuan Chen","doi":"10.1016/j.pnucene.2026.106241","DOIUrl":"10.1016/j.pnucene.2026.106241","url":null,"abstract":"<div><div>The nuclear safety video display unit (SVDU) is a critical human-machine interface in the nuclear reactor protection system, and its cybersecurity is increasingly challenged by the sector's full digitalization. This paper proposes a novel cybersecurity risk assessment method integrating an attack tree model with the STRIDE threat framework to evaluate the risk status of the nuclear SVDU. The approach innovatively incorporates the latest CVSS4.0 standard to define security attributes (attack cost, technical difficulty, discovery difficulty, and impact) for leaf nodes, which are quantified through hierarchical scoring. Crucially, an objective entropy weighting method is employed to calculate the weights of these attributes, effectively eliminating the subjective bias inherent in traditional expert-dependent methods like FAHP. The risk probability is then propagated from leaf to root nodes based on their dependency relationships. Comparative analysis demonstrates that the proposed method offers a more sensitive and discriminating assessment, identifying critical threats that subjective approaches might overlook and providing a robust, data-driven basis for prioritizing security protection strategies.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106241"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145927222","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 the “double peaks” heat transfer characteristics of supercritical carbon dioxide","authors":"Xiaoyang Lun ,&nbsp;Haicai Lyu ,&nbsp;Xianjun Chen ,&nbsp;Jinghui Wu ,&nbsp;Cailu Xu ,&nbsp;Mingyun Liu ,&nbsp;Fenglei Niu ,&nbsp;Fang Liu ,&nbsp;Han Wang ,&nbsp;Yanping Huang","doi":"10.1016/j.pnucene.2025.106204","DOIUrl":"10.1016/j.pnucene.2025.106204","url":null,"abstract":"<div><div>This paper presents an experimental investigation of supercritical carbon dioxide (sCO₂) flow in a vertically oriented circular channel with an inner diameter of 4 mm. The study systematically examined the effects of heat flux, mass flux, and pressure on heat transfer characteristics within parameter ranges of 214.7–502.7 kg m⁻²·s⁻¹ for mass flux, 8.0–9.2 MPa for pressure, and 33.4–68.8 kW m⁻² for heat flux. Under multiple operational conditions, a double-peak wall temperature distribution was observed, for which a conceptual definition was established. The formation mechanism of this phenomenon was discussed through the framework of buoyancy-driven effects based on single-phase fluid theory. The results demonstrate that the buoyancy criterion <em>Bu</em> provides superior prediction for the onset of heat transfer deterioration. Through experimental data analysis, the threshold value of <em>Bu</em> was determined to be 5.0 × 10⁻⁴. Further, the buoyancy criterion <em>Bu</em> is proposed to develop new predicted Nusselt numbers for the double peaks heat transfer in supercritical pressures, which is significance for the engineering application of sCO₂ power cycles.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106204"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145884667","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":"Development and validation of a pin-by-pin neutronics/thermal-hydraulics transient coupling code for PWRs based on the SP3 method","authors":"Zhigang Li ,&nbsp;Wenbo Zhao ,&nbsp;Bangyang Xia ,&nbsp;Xiaoming Chai ,&nbsp;Wei Zeng ,&nbsp;Junjie Pan ,&nbsp;Xiafeng Zhou ,&nbsp;Shenglong Qiang ,&nbsp;Wei Lu","doi":"10.1016/j.pnucene.2025.106194","DOIUrl":"10.1016/j.pnucene.2025.106194","url":null,"abstract":"<div><div>Solving the neutron transport equation by using the SP_n method to achieve full-core pin-by-pin calculation represents a significant research direction in nuclear reactor numerical simulation. This work builds upon CORCA-SPn's existing capability to solve the steady-state SP₃ equations using the Two-Node Method combined with Coarse Mesh Finite Difference (CMFD) and Fine Mesh Finite Difference (FMFD). This solution methodology was extended to solve the 3-dimensional(3D) SP₃ space-time kinetics equations in this paper. The backward Euler scheme and theta method are respectively used for the discretization of neutron dynamics SP₃ equations and fuel thermal conductivity equations in CORCA-SPn. Furthermore, leveraging the Multi-physics Object-oriented Reactor Engineering (MORE) framework developed by the Nuclear Power Institute of China (NPIC), the newly developed SP₃ transient transport module is coupled with the existing single-channel thermal-hydraulics module TH1D. This integration enables transient full-core pin-by-pin neutronics/thermal-hydraulics(N-TH) coupling calculation code. The Picard method and the operator splitting(OS) method are respectively employed to govern the iterative procedures for steady-state and transient N-TH coupling calculations in CORCA-SPn. The code's capability for solving SP₃ space-time kinetics problems was validated by using C5G7-TD benchmark. Its transient N-TH coupling capability was validated by using PWR MOX/UO₂ benchmark. The results demonstrate that: 1) CORCA-SPn exhibits good accuracy in solving SP₃ space-time kinetics equations and performing transient N-TH coupled calculations, reliably predicting key transient parameters such as core relative power, reactivity, and fuel Doppler temperature. 2) Validation against the neutron kinetics benchmark of C5G7-TD shows a maximum relative deviation in transient peak power of 6.41 % compared to MPACT. 3) Validation against the control rod ejection benchmark of PWR MOX/UO₂ shows the maximum relative deviation in transient peak power is 20.1 %, the maximum deviation in peak reactivity is −0.05 $, the maximum deviation in fuel Doppler temperature is 3.39 °C, and the maximum deviation in average coolant temperature is 0.19 °C compared to PARCS and Serpent/SCF.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106194"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145788989","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":"Neutronic study of ThF4-UF4-LiF fuel mixture in the molten salt hybrid reactor for 233U denaturing","authors":"Hacı Mehmet Şahin ,&nbsp;Sümer Şahin ,&nbsp;Güven Tunç ,&nbsp;Hüseyin Şahiner","doi":"10.1016/j.pnucene.2026.106237","DOIUrl":"10.1016/j.pnucene.2026.106237","url":null,"abstract":"<div><div>This study investigates a fusion-fission hybrid reactor concept using a thorium-based molten salt fuel mixture to enhance proliferation resistance and operational sustainability. Neutron transport and reaction rates were modeled using the Monte Carlo N-particle code (MCNP6) with ENDF/B-VIII.0 data. Thorium is mixed homogeneously with 2.25 % depleted uranium (DU) in order to denaturate the ²³³U fuel. The analysis. showed that with 75 % ⁶Li enrichment and a 50 cm coolant layer, the tritium breeding ratio (TBR) remained above 1.05 for a period of four years. The energy multiplication factor (M) increased from 1.88 to 2.2, consistently exceeding the minimum target of 1.5. Under the hard fusion neutron flux, more than 96 % of the plutonium produced was ²³⁹Pu, heavier plutonium isotopes were burnt <em>in situ</em>. The production of the low enriched ²³³U fuel increased to about 12 % after 34 months. These results indicate the technical feasibility of a thorium-based fusion-fission hybrid reactor with improved proliferation resistance, efficient energy multiplication, and sustainable fuel cycle characteristics.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106237"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145927144","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":"PuO2 surrogate behaviour in a nuclear waste incineration process","authors":"A. Quintas ,&nbsp;P. Charvin ,&nbsp;S. Lemonnier ,&nbsp;M. Tribet ,&nbsp;A. Russello ,&nbsp;K. Poizot ,&nbsp;H. Pablo ,&nbsp;B. Frasca","doi":"10.1016/j.pnucene.2025.106229","DOIUrl":"10.1016/j.pnucene.2025.106229","url":null,"abstract":"<div><div>In this study, the behaviour and fate of plutonium oxide within a nuclear waste incineration reactor was investigated. A comprehensive literature review on the use of PuO₂ surrogate in the nuclear R&amp;D sector was first undertaken to assess different scientific approaches and related limitations. A thorough analysis of the different phenomenologies involved in the incineration process was also performed which resulted in the selection of a set of four oxide surrogates (ZrO₂, CeO₂, Gd₂O₃ and HfO₂) that were eventually used during seven inactive full-scale mock-up test campaigns. After description of the innovative methodology and the experimental set-up, the results obtained were presented. Surrogate retention factors were calculated for a wide range of operating conditions and systematically stands below 1 %. The use of several surrogate oxide powders with different particle size distributions ranging from 0.05 to 120 μm allowed to put in evidence interesting granulometric effects which were supported by some preliminary numerical simulation studies. It has also been confirmed that the fiberglass bag used to wrap the waste load in the incineration reactor plays a key role in the containment of surrogate powders. The data and observations were finally consistent with the existence of a dynamic deposition process on the reactor cooled walls that combines growth and elimination stages. This mechanism requires to be confirmed by further investigations over longer period of time and could open the way to a steady state process operation with a constant equilibrium mass of actinide trapped within the reactor deposit.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106229"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145927226","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 simulation of the flow flashing of carbon dioxide in a circular pipe","authors":"Wukun Zhu,&nbsp;Han Wang,&nbsp;Fanghao Chen,&nbsp;Jinghui Wu","doi":"10.1016/j.pnucene.2025.106224","DOIUrl":"10.1016/j.pnucene.2025.106224","url":null,"abstract":"<div><div>The study of Loss of Coolant Accidents (LOCA) in supercritical carbon dioxide (S-CO₂) reactors constitutes a critical and challenging domain in nuclear safety analysis, with the investigation of flow flashing phenomena in pipelines being particularly pivotal. In this work, reactor pipelines were simplified, and a transient numerical simulation code for channel flow flashing was developed based on the homogeneous flow model, integrated with a suite of relevant correlations. Computational results from this code were validated against experimental data for water flow flashing, showing strong consistency with the measured values. Specifically, the maximum error in predicting flash onset (the point where vapor first forms) was approximately 5 %, thereby confirming the model accuracy. Utilizing the validated code, an in-depth analysis of the flow flashing process with CO₂ as the working fluid was conducted. For vertical pipes, higher inlet fluid temperature, lower outlet pressure, and lower inlet velocity were found to extend the length of the flashing two-phase region and increase the outlet void fraction. Furthermore, transient depressurization processes at the outlets of vertical pipes were numerically simulated to verify the code's transient analysis capabilities. For horizontal pipes, lower inlet pressure triggers adiabatic flashing, resulting in higher steam quality. Faster outlet depressurization initiates earlier flashing, enhances volume expansion, accelerates flow velocity, and elevates the peak mass flow rate. Overall, the computational results provide a solid theoretical foundation for LOCA analysis in S-CO₂ reactors.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106224"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145884669","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}