Nuclear Engineering and Design最新文献

{"title":"An adaptive statistical model of nuclear rods temperature for the detection of total and instantaneous blockage","authors":"Rémi Cogranne","doi":"10.1016/j.nucengdes.2025.114243","DOIUrl":"10.1016/j.nucengdes.2025.114243","url":null,"abstract":"<div><div>The monitoring of critical systems is of the utmost importance, especially when undetected malfunctions could lead to major accidents. This paper focuses on the temperature monitoring of fuel rod assemblies within nuclear power plants, with the goal of detecting total and instantaneous blockages as reliably and quickly as possible. First, we address the modelling of the temperature of the whole fuel rod assembly altogether. We propose a linear parametric model that is adaptive, incorporating previous temperature measurements to enhance its accuracy. This approach allows us to distinguish between regular, non-anomalous temperatures and the anomalous thermal event due to a blockage. The proposed sequential, or online, detection scheme is reliable, as the false alarm rate and detection power are analytically bounded. The model and subsequent statistical test are generic, making the methodology applicable to a wide range of nuclear cores. Numerical experiments, using real temperature measurements from the Superphénix power station, demonstrate the accuracy of the proposed model and the relevance of the detection procedure.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"442 ","pages":"Article 114243"},"PeriodicalIF":1.9,"publicationDate":"2025-06-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144502319","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":"Open-source microreactor design models for technoeconomic assessments","authors":"Khaldoon Al-Dawood ,&nbsp;Botros Hanna ,&nbsp;Sai P. Balla ,&nbsp;Rodrigo de Oliveira ,&nbsp;Sam Garcia ,&nbsp;Dan McCarthy ,&nbsp;Chandu Bolisetti ,&nbsp;Ben Lindley ,&nbsp;Abdalla Abou-Jaoude","doi":"10.1016/j.nucengdes.2025.114210","DOIUrl":"10.1016/j.nucengdes.2025.114210","url":null,"abstract":"<div><div>Technoeconomic analyses for advanced nuclear technologies are essential for identifying cost drivers which allow for optimizing the technology to achieve better economic performance. Microreactors are emerging as a promising and reliable-energy solution, offering inherent safety, low capital investment, and rapid deployment capabilities. Past studies have conducted technoeconomic analyses for microreactors, but there remains a need for open-source technoeconomic models that can enhance collaboration, transparency, and consistency when performing this type of analysis. The present article introduces an open-source technoeconomic model for microreactors based on bottom-up cost estimation. Since no real cost data for microreactors exists, this model leverages cost data and insights gleaned from the Microreactor Applications, Research, Validation and Evaluation (MARVEL) project. It estimates the first-of-a-kind cost of two microreactor technologies and can also calculate the N^th of a kind (NOAK) cost via accounting for learning and mass factory production. The two technologies considered in this paper are the liquid–metal thermal reactor (LTMR) and the gas-cooled microreactor (GCMR). The goal of this study is to demonstrate bottom-up cost estimation of these microreactor technologies and provide the cost estimation models that other users can leverage for various applications such as design optimization and financial planning.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"442 ","pages":"Article 114210"},"PeriodicalIF":1.9,"publicationDate":"2025-06-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144502320","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":"High-resolution predictions of the coolant properties for the 3D PWR core with artificial neural networks based on CTF","authors":"Marianna Papadionysiou ,&nbsp;Gregory Delipei ,&nbsp;Maria Avramova ,&nbsp;Hakim Ferroukhi ,&nbsp;Kostadin Ivanov","doi":"10.1016/j.nucengdes.2025.114261","DOIUrl":"10.1016/j.nucengdes.2025.114261","url":null,"abstract":"<div><div>PSI and North Carolina State University are developing a high-resolution multi-physics core solver for Pressurized Water Reactor (PWR) analysis in Cartesian geometry, using the neutron transport code nTRACER and two Machine Learning (ML) models providing thermal–hydraulic (T/H) feedback. This work focuses on the ML models, trained with CTF data to predict PWR subchannel coolant properties during normal operation. The methodology presented can be applied to different PWR core, to produce ML models capable of high-resolution T/H predictions. The ML model’s performance is evaluated on quarter-core CTF calculations achieving an average temperature difference of 1 °C from CTF and equivalent density accuracy. Their verification is extended outside configurations seen in their training, with varying mass flux, a water liner and different lattice geometry. They are also compared to a simplified one-dimensional T/H solver, showing significantly lower discrepancies, almost half, with similar computational cost, while being up to 15 times faster than CTF.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"442 ","pages":"Article 114261"},"PeriodicalIF":1.9,"publicationDate":"2025-06-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144502318","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 investigation of the internal structure of boiling two-phase water flow under LWR core operating conditions","authors":"J.-M. Le Corre ,&nbsp;H. Li ,&nbsp;D. Grishchenko ,&nbsp;A. Kipiela ,&nbsp;M. Persson ,&nbsp;P. Kudinov ,&nbsp;H. Anglart","doi":"10.1016/j.nucengdes.2025.114249","DOIUrl":"10.1016/j.nucengdes.2025.114249","url":null,"abstract":"<div><div>An experimental setup has been designed and manufactured at the Royal Institute of Technology (KTH) in Stockholm to investigate the internal structure of boiling two-phase water flow under prototypical Light Water Reactor (LWR) core conditions, including those relevant to PWR, BWR and SMR designs. The setup is based on the High-pressure WAter Test (HWAT) loop, designed for 25 MPa pressure, 1 kg/s water mass flow rate and 1 MW thermal power. The facility has been updated with a new test section and advanced instrumentation systems to enable measurements in both forced convection and natural circulation, under steady-state and transient operations. This novel experimental setup allows for the first-time measurements of radial distributions of local two-phase flow parameters under high-pressure LWR core conditions. The resulting data is intended to enhance the fundamental understanding of boiling two-phase flow phenomena, contribute to the development of closure laws (including for polydispersed flow) and support the validation of computational codes (1-D and 3-D). The paper presents the loop design, the updated instrumentation with associated uncertainties, and data post-processing methods (including the derivation of dispersed phase length scales). Results from commissioning tests, such as heat balance tests and single-phase tests, are presented. Examples of high-pressure boiling two-phase flow measurements are presented and discussed. Fundamental behavior and associated key parameters, including radial distributions of void fraction, mixture velocity, interfacial length scales and polydispersed characteristics, are identified and quantified.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"442 ","pages":"Article 114249"},"PeriodicalIF":1.9,"publicationDate":"2025-06-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144502316","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 investigation on the irradiation-induced mechanics behavior in U3Si2/Al dispersion fuel using RVE model","authors":"Wenwen Zhang ,&nbsp;Wenli Guo ,&nbsp;Wentao Hao ,&nbsp;Ding She ,&nbsp;Wei Li","doi":"10.1016/j.nucengdes.2025.114270","DOIUrl":"10.1016/j.nucengdes.2025.114270","url":null,"abstract":"<div><div>A representative volume element (RVE) model was developed to investigate the thermo-mechanical behavior of U₃Si₂/Al dispersion fuel under irradiation. The study focused on the evolution of thermal expansion, elastic properties, creep, and irradiation-induced swelling across varying irradiation conditions. RVE models with different uranium densities were constructed using Voronoi tessellation to approximate the irregular geometry of fuel particles. A phase-field approach was employed to simulate the growth of the interaction layer. Coupled with the finite element method (FEM), a series of detailed simulations were conducted. The simulations revealed that increasing interaction layer thickness (from 0 μm to 20 μm) significantly reduces the thermal expansion coefficient by up to 28 % while increasing stiffness. Higher fission density and fuel particle volume fraction lead to greater swelling strain. The effect of interaction layer thickness on equivalent swelling strain exhibits a dual behavior. The creep behavior is affected by both the interaction layer volume fraction and the applied stress, and can be effectively described by a polynomial-based constitutive model. Compared to traditional homogenization methods, the RVE approach offers improved accuracy in capturing the effects of microstructural heterogeneity on macroscopic fuel performance. These results demonstrate the model applicability for evaluating and optimizing dispersion fuel behavior under irradiation.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"442 ","pages":"Article 114270"},"PeriodicalIF":1.9,"publicationDate":"2025-06-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144502317","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":"Innovative thorium-based fuel assemblies for LW-SMR: In-depth assembly-level neutronic analysis and safety considerations in solid and annular configurations","authors":"Hanan Rifai ,&nbsp;Ouadie Kabach ,&nbsp;Zouhair Sadoune ,&nbsp;El Mahjoub Chakir ,&nbsp;Sinem Uzun ,&nbsp;Hamid Amsil ,&nbsp;Fadi El Banni","doi":"10.1016/j.nucengdes.2025.114274","DOIUrl":"10.1016/j.nucengdes.2025.114274","url":null,"abstract":"<div><div>This study presents a comprehensive neutronic and safety evaluation of different thorium-based fuel compositions (Th-²³⁵U)O₂, (Th-²³³U)O₂, and a new proposed composition, namely, (Th-²³³U-²³⁵U)O₂ fuel, compared to conventional UO₂ in solid and dual-cooled annular assembly configurations in the context of Small Modular Reactors (SMR) applications. Results demonstrate that (Th-²³³U)O₂ achieves the highest cycle length in both solid configurations, significantly exceeding UO₂ performance. The proposed, however, (Th-²³³U-²³⁵U)O₂ composition exhibits favorable safety characteristics, including enhanced negative temperature reactivity coefficients and favorable kinetic parameters. Annular fuel configurations also demonstrate superior thermal performance with centerline temperature reductions of approximately 50 °C compared to solid configurations and consistently lower power peaking factors throughout burnup. The findings indicate that thorium-based fuels, particularly (Th-²³³U-²³⁵U)O₂ in dual-cooled annular configuration, offer compelling advantages for SMR applications, including extended fuel cycles, improved safety characteristics, enhanced proliferation resistance, and superior thermal performance, supporting their viability as sustainable alternatives to conventional uranium-based fuels.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"442 ","pages":"Article 114274"},"PeriodicalIF":1.9,"publicationDate":"2025-06-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144489845","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 optimization of a closed-jet trans-critical hydrogen moderator with non-uniform heat sources using machine learning and multi-objective algorithms","authors":"Jiahui Chen ,&nbsp;Jianfei Tong ,&nbsp;Youlian Lu ,&nbsp;Chaoju Yu ,&nbsp;Yu Zhen ,&nbsp;Songlin Wang ,&nbsp;Bin Zhou ,&nbsp;Congju Yao ,&nbsp;Tianjiao Liang ,&nbsp;Wen Yin ,&nbsp;Jian Wen","doi":"10.1016/j.nucengdes.2025.114242","DOIUrl":"10.1016/j.nucengdes.2025.114242","url":null,"abstract":"<div><div>The optimization of heat transfer in transcritical hydrogen moderators, such as the Decoupled Poisoned Hydrogen Moderator (DPHM) at the China Spallation Neutron Source (CSNS), is critical for enhancing neutron flux and ensuring the safe operation of spallation neutron sources. This study addresses the unique thermal–hydraulic challenges posed by closed-jet impingement and non-uniform internal heat sources in transcritical hydrogen systems, which are essential for maintaining efficient neutron moderation and thermal stability. A three-dimensional computational model is developed to investigate the effects of key design parameters, including the flow baffle offset (<em>L</em>), inlet jet height (<em>H</em>), and hydrogen mass flow rate (<em>m</em>), on heat transfer and flow dynamics. Leveraging a genetic algorithm-optimized backpropagation (BP) neural network, this work introduces an innovative predictive framework to capture the complex nonlinear relationships between these parameters and system performance. Furthermore, a multi-objective optimization approach combining NSGA-III and TOPSIS is employed to identify optimal operating conditions that balance cooling efficiency and flow velocity constraints. The results demonstrate the critical role of jet impingement in mitigating thermal stratification and improving heat transfer, offering significant insights for the design and optimization of hydrogen moderators in advanced nuclear systems. This study not only advances the understanding of transcritical hydrogen behavior under non-uniform heat sources but also provides a practical, machine learning-enhanced methodology for optimizing thermal–hydraulic performance in spallation neutron sources.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"442 ","pages":"Article 114242"},"PeriodicalIF":1.9,"publicationDate":"2025-06-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144480838","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":"Feasibility of small modular reactors for decarbonizing district heating systems: a case study of the Helsinki metropolitan area","authors":"Jussi-Pekka Ikonen,&nbsp;Tomi J. Lindroos,&nbsp;Pauli Hiltunen","doi":"10.1016/j.nucengdes.2025.114262","DOIUrl":"10.1016/j.nucengdes.2025.114262","url":null,"abstract":"<div><div>This study explores the feasibility of integrating small modular reactors (SMRs) into the district heating (DH) and cooling system of the Helsinki metropolitan area, Finland. The objectives of the study are to evaluate the economic and operational aspects of two SMR technologies, the academic concept E-SMR (155 MWelectric/50 MW_DH) and the VTT-developed LDR-50 (50 MW_DH), as well as their environmental impact on the DH and cooling system using the Backbone modeling framework. Despite investments in biomass-fired DH generation and large heat pumps, Helsinki remains heavily dependent on fossil fuels. SMRs could potentially offer a reliable and continuous supply of carbon-free energy.</div><div>Results indicate that while both SMR designs reduce CO2 emissions and enhance electricity supply, E-SMR is not a profitable investment, whereas LDR-50 is. The fluctuating energy demand and low electricity prices in the studied area presented challenges for the larger E-SMR. Heat production capabilities are highly valued in the studied system, which was reflected in LDR-50′s profitability and the sensitivity study of E-SMR’s heat extraction rate.</div><div>Sensitivity analyses highlighted the influence of economic factors such as investment costs, discount rates, and operating expenses on the profitability of both SMRs. Additionally, energy market conditions play a crucial role in determining the financial viability of SMRs, suggesting that their profitability can vary significantly under different market scenarios. Integrating SMRs into future energy systems, such as electrolyzers, could enhance their feasibility, warranting further research. Moreover, the economic viability of different fuel strategies should be investigated to explore their potential benefits for SMRs.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"442 ","pages":"Article 114262"},"PeriodicalIF":1.9,"publicationDate":"2025-06-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144489844","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":"Photonuclear treatment for spent fuel radiotoxicity reduction: a case study investigation on minor actinides","authors":"Antonio Cammi ,&nbsp;Lorenzo Loi ,&nbsp;Andrea Missaglia ,&nbsp;Ludovica Tumminelli ,&nbsp;Francesca Giacobbo ,&nbsp;Enrico Padovani","doi":"10.1016/j.nucengdes.2025.114204","DOIUrl":"10.1016/j.nucengdes.2025.114204","url":null,"abstract":"<div><div>The management of Spent Nuclear Fuel (SNF) is one of the main challenges in the decommissioning of nuclear power plants. Thermal reactors, such as Light Water Reactors (LWRs), produce significant amounts of minor actinides (MAs) such as Americium, Curium, and Neptunium, which are key contributors to the long-term radiotoxicity and decay heat in SNF. Currently, the long term widely accepted solution is the geological disposal. At the same time, advanced technologies like Partitioning and Transmutation (P&amp;T) offer promising solutions to reduce SNF long-term radiotoxicity. While most transmutation strategies rely on neutron fluxes, in this study the adoption of photon beam to induce photonuclear reactions in SNF is investigated, without depending on neutron based systems. In particular, the study focuses on the probability of inducing transmutations and fissions on MAs, by leveraging the Giant Dipole Resonance (GDR) region of photonuclear interactions. In the presented case study, the aim was to investigate the effect of a photon driven transmutation of minor actinides present in a spent fuel from SMR technology. The focus was mainly put on studying the physics of the system, analysing the feasibility of reducing the inventory and radiotoxicity of the system by this method, without considering technological aspects and limitations.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"442 ","pages":"Article 114204"},"PeriodicalIF":1.9,"publicationDate":"2025-06-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144470199","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":"The tail-oriented multi-normal model method for partially correlated seismic fragilities in probabilistic risk assessment","authors":"Mohamed M. Talaat ,&nbsp;Abhinav Anup","doi":"10.1016/j.nucengdes.2025.114217","DOIUrl":"10.1016/j.nucengdes.2025.114217","url":null,"abstract":"&lt;div&gt;&lt;div&gt;Many seismic-induced failures are due to shaking effects in response to common ground motions. Since the effects of shaking are propagated to equipment with varying configurations located at the same or different locations, there is often a degree of partial correlation between these failures. Common practice in seismic probabilistic risk assessment (SPRA) of nuclear facilities typically idealizes this partial correlation as either perfect correlation or full independence. Recent studies indicate that refining this practice for structures or components with considerable risk contributions can lead to significant effects on the SPRA outcome. Methods and techniques to incorporate the modelling of partial correlation in SPRA models have received increasing research attention in recent years, and a few recent SPRAs explicitly modelled it for risk-significant contributors. While applicable to single-unit (SU) SPRA, explicit modelling of seismic fragility partial correlation may be even more impactful for multi-unit (MU) SPRA.&lt;/div&gt;&lt;div&gt;Several methods with various degrees of rigor exist in the literature for explicitly modelling partial correlation between seismic fragilities in a probabilistic risk quantification. The available rigorous methods typically involve the use of numerical simulation or computationally intensive numerical analysis, which are not practical for inclusion in SPRA qualification. This article discusses the tail-oriented multi-normal model (TMM) method, a robust technique for explicitly modelling partial correlation between seismic fragilities in SPRA applications. The TMM method is based on the Separation of Independent and Common Variables (SICV) concept and is superior to other SICV-based techniques in that it uses an efficient closed-form formulation. This analytical solution is premised on the lognormal probability distribution being a valid representation of seismic-induced failure fragilities, which is commonly accepted in SPRA.&lt;/div&gt;&lt;div&gt;This article is divided into an introduction and four sections. The introduction reviews available relevant literature on the topic, presents relevant technical background elements, and summarizes specific current technical challenges. The first section introduces the TMM method formulation, and implementation steps. The second section presents validation examples. The third section discusses the determination of partial correlation factors and the sensitivity of the TMM method results to the accuracy of the partial correlation factors. The fourth section discusses the use of the TMM method in SPRA models.&lt;/div&gt;&lt;div&gt;The performance of the TMM method is found to be more favourable than the widely used Reed-McCann method both computationally (processing time) and analytically (solution accuracy). The TMM method results are found to be weakly sensitive to fragility analyst judgement in quantifying partial correlation, making it robust against potential bias in such judgements. Furt","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"442 ","pages":"Article 114217"},"PeriodicalIF":1.9,"publicationDate":"2025-06-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144470666","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}