Annals of Nuclear Energy最新文献

{"title":"Systematic generation of effective thermophysical and thermo-mechanical properties for reduced-order modeling of TRISO fuel pellets","authors":"Samuel Heflin ,&nbsp;Vedant Mehta ,&nbsp;Dan Kotlyar","doi":"10.1016/j.anucene.2025.111487","DOIUrl":"10.1016/j.anucene.2025.111487","url":null,"abstract":"<div><div>This work presents a detailed methodology for generating effective material properties of tri-structural isotropic (TRISO) fuel pellets for use in reduced-order homogenized thermal and/or mechanical calculations. This is achieved by use of the finite-element (FE) method to perform simulations of explicitly modeled TRISO fuel pellets, in conjunction with derivation of analytical equations for analogous homogenous pellets which preserve a particular parameter of interest. This method is applied to several properties including thermal conductivity, specific heat capacity, thermal expansion coefficient, Young’s Modulus and Poisson’s Ratio. The properties generated are a function of temperature and packing fraction. It is demonstrated that the state-of-the-art for homogenizing thermal conductivity may be inadequate for realistic TRISO pellets, while for other properties existing homogenization standards are scarce in the literature. Finally, a multi-physics application of the reduced-order properties is demonstrated, which involves coupling neutronics with thermo-mechanical calculations. The fission power distribution is obtained from neutronic analysis, and then temperature, displacement and stress–strain distributions from the explicit finite-element model are compared with those from the homogenous calculations. For the finite-element simulations, the commercial software Abaqus (<span><span>ABAQUS, 2014</span></span>) is used, and the Monte Carlo code MCNP is used for neutronics to obtain the power distributions.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"219 ","pages":"Article 111487"},"PeriodicalIF":1.9,"publicationDate":"2025-04-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143833913","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":"Improvement of in-vessel corium transient stratification modelling regarding the uranium and zirconium phase partitioning","authors":"J. Mativet,&nbsp;A. Lecoanet,&nbsp;R. Le Tellier","doi":"10.1016/j.anucene.2025.111444","DOIUrl":"10.1016/j.anucene.2025.111444","url":null,"abstract":"<div><div>One of the main threats on In-Vessel Melt Retention (IVR) mitigation strategy is the transient thermochemical stratification of the liquid corium in the vessel lowerhead. This paper proposes to improve the modelling of the transient liquid phase stratification of the {U,O,Zr,Steel} system within its miscibility gap. This work is an increment on a previous model considering the metal droplet forming during the relocation, and the focus is here on improving the mass transfer modelling by treating Uranium and Zirconium separately and not as a pseudo component. This means that in the model presented here the molar ratio of Uranium over Zirconium is not equal in the oxide and in the metallic phase.</div><div>The equations of the model are presented as well as the thermodynamic closures it requires. Dissociating the transfers of Uranium and Zirconium allows for the use of thermodynamic closures that are closer to the actual thermodynamic state calculated using a Gibbs Energy minimizer.</div><div>The improved model is applied to a synthetic transient where metal is added on top of an oxide phase leading to an equilibrium state with heavy metal layer creation. This synthetic transient is further modified to reach stratification inversion. The results are compared to the previous model and show significant improvements in computing the transient stratification.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"219 ","pages":"Article 111444"},"PeriodicalIF":1.9,"publicationDate":"2025-04-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143828625","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":"Research on sensor anomaly detection and fault location in nuclear power plants based on GAN-IAAKR model","authors":"Jiarong Gao,&nbsp;Yongkuo Liu,&nbsp;Qiang Zhao,&nbsp;Xin Ai,&nbsp;Longfei Shan","doi":"10.1016/j.anucene.2025.111462","DOIUrl":"10.1016/j.anucene.2025.111462","url":null,"abstract":"<div><div>To address the issues of accurate anomaly detection under minor drift and accuracy degradation faults in nuclear power plant sensors, as well as the lack of data directionality, model robustness, and real-time performance in fault localization algorithms for nuclear power sensors. A GAN-IAAKR-based method for sensor anomaly detection and fault localization is proposed in this paper to accurately and promptly detect anomalous states in nuclear power plant sensors during faults and to perform fault localization for the faulty sensors First, a time-series two-dimensionalization method is employed to encode normal sensor monitoring data into images. Next, evolutionary game theory is utilized, where multiple generators and discriminators are organized into two distinct groups to simultaneously engage in the game. During the game, a replicator dynamic equation is established, and the weights of each generator and discriminator are dynamically adjusted through a smoothing mechanism. The encoded images are then fed into an enhanced GAN model for training. Building on this, the GAN model reconstructs the images, and anomaly detection is carried out by calculating the MSE statistic (reconstruction error) between the original and the reconstructed images. Finally, the improved AAKR algorithm is applied to reconstruct the sensor monitoring data. By setting a reconstruction error threshold, normal and anomalous data can be effectively differentiated, thereby enabling fault localization. Experiments were conducted to compare anomaly detection performance between One-Class SVM, Autoencoder, and GAN models under conditions of minor drift and accuracy degradation faults. The experimental results demonstrate that the proposed anomaly detection model exhibits a high detection accuracy, exceeding 95%, while the fault localization models also show superior accuracy, real-time performance, and robustness, with an accuracy rate exceeding 90%.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"219 ","pages":""},"PeriodicalIF":1.9,"publicationDate":"2025-04-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143828627","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 CoRREx neutron spectrum filtering campaign at VENUS-F for calculation-to-experiment discrepancy interpretation","authors":"Federico Grimaldi ,&nbsp;Federico Di Croce ,&nbsp;Antonin Krása ,&nbsp;Grégoire de Izarra ,&nbsp;Loic Barbot ,&nbsp;Patrick Blaise ,&nbsp;Pierre-Etienne Labeau ,&nbsp;Luca Fiorito ,&nbsp;Guido Vittiglio ,&nbsp;Jan Wagemans","doi":"10.1016/j.anucene.2025.111425","DOIUrl":"10.1016/j.anucene.2025.111425","url":null,"abstract":"<div><div>Research reactors and particularly zero power reactors serve as bridges between reactor design and deployment. The fast spectrum VENUS-F zero power reactor operated at SCK CEN has been conceived to support the development of fast heavy-metal-cooled reactor designs. Over time, several discrepancies between model results and experiments were observed, especially where the neutron spectrum is more epithermal. To identify their origin, the CoRREx experiment was conducted: B<span><math><msub><mrow></mrow><mrow><mn>4</mn></mrow></msub></math></span>C filters of different thickness were loaded in VENUS-F to investigate the relevance of the epithermal neutron spectrum to those discrepancies. We present measurements of spectral indices (fission rate ratios), <span><math><msup><mrow></mrow><mrow><mn>235</mn></mrow></msup></math></span>U fission rate traverses and reactivity worth and compare them with the results of Serpent calculations using JEFF-3.3, ENDF/B-VIII.0, ENDF/B-VIII.1 and JENDL-4.0u. CoRREx shows that the Serpent model performance in reproducing the experiment results is rather similar in the active core center and periphery. Furthermore, using B<span><math><msub><mrow></mrow><mrow><mn>4</mn></mrow></msub></math></span>C neutron spectrum filters and conducting a thorough propagation of experimental uncertainty CoRREx provides insights on the minimal contribution of the epithermal neutron flux spectrum component to the observed discrepancies. CoRREx indicates that the observed spectral index calculation-to-experiment discrepancy is attributable to the fast neutron spectrum. The Serpent model shows good performance for <span><math><msup><mrow></mrow><mrow><mn>235</mn></mrow></msup></math></span>U fission rate traverses and a slight underestimation of the filter reactivity worth.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"219 ","pages":"Article 111425"},"PeriodicalIF":1.9,"publicationDate":"2025-04-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143828628","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 light gas transport during containment venting by using the large-scale test facility CIGMA","authors":"Shu Soma ,&nbsp;Masahiro Ishigaki ,&nbsp;Yasuteru Sibamoto","doi":"10.1016/j.anucene.2025.111455","DOIUrl":"10.1016/j.anucene.2025.111455","url":null,"abstract":"<div><div>Containment venting is one of the accident mitigation measures during severe accidents in nuclear power plants for preventing overpressure failure of the containment vessels. Because of the capability of releasing hydrogen generated in the containment vessel, the hydrogen risk can be also reduced. In this study, we conducted experiments with the large-scale test facility CIGMA to investigate the light gas transport during the venting action, mainly focusing on the effect of sump water boiling caused by the vent. The CIGMA test vessel initially pressurized by steam, air, and helium (hydrogen simulant) that formed a helium-rich density stratification was depressurized with and without sump water, with different venting flow rates, and at different venting positions. As the sump water became a steam source due to flash boiling, the helium stratification was diluted and the venting time increased twofold compared to the case without sump water, which significantly affected the amount of helium discharged to the atmosphere. Especially for the high venting flow rate condition, the amount of helium remaining in the vessel at the end of depressurization was half that of the case without sump water. Lowering the venting position from within the initial stratification to 3 m below its interface led to a threefold increase in the amount of helium remaining at the same low pressure, because of the longer time until the helium-rich stratification reached the venting position.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"219 ","pages":"Article 111455"},"PeriodicalIF":1.9,"publicationDate":"2025-04-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143833912","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 new mathematical model for the radionuclide concentrations in crops for the generic biosphere assessment","authors":"Dongki Kim,&nbsp;Wontak Lee,&nbsp;Jonghyun Kim,&nbsp;Jongtae Jeong,&nbsp;Soonyoung Kim,&nbsp;Joowan Park","doi":"10.1016/j.anucene.2025.111447","DOIUrl":"10.1016/j.anucene.2025.111447","url":null,"abstract":"<div><div>A new mathematical model for radionuclide concentrations in crops is developed to support a generic safety assessment for a deep geological repository in Korea. It is developed by reflecting domestic agricultural practices such as land use, harvest and fertilization, irrigation method, and food processing. A new approach is applied to the external contaminants on crop surfaces during two types of periods such as the irrigation period and the period from the end of the irrigation period to the harvest. Considering the significant impacts of radionuclide concentrations in crops to those in animal products and total exposure doses for the well water farming scenario, the mathematical model developed can support a more realistic safety assessment and the development of a safety case for the deep geological repository. In addition, the separated weathering duration approach adopted is expected to provide flexibility for the changes in the irrigation period considering climate change.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"219 ","pages":"Article 111447"},"PeriodicalIF":1.9,"publicationDate":"2025-04-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143828626","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":"Impact of U-10Mo HALEU fuel element tolerances on the Massachusetts Institute of Technology reactor safety and operational performance – Thermal hydraulics","authors":"Sero Yang ,&nbsp;Valerio Mascolino ,&nbsp;Laura M. Jamison ,&nbsp;Kyle E. Anderson ,&nbsp;Lin-wen Hu ,&nbsp;Dhongik S. Yoon ,&nbsp;John A. Stillman ,&nbsp;Walid M. Mohamed ,&nbsp;Erik H. Wilson","doi":"10.1016/j.anucene.2025.111376","DOIUrl":"10.1016/j.anucene.2025.111376","url":null,"abstract":"<div><div>The U.S. is coordinating efforts for the conversion of six U.S. High Performance Research Reactors (USHPRR) including one critical facility from highly enriched uranium (HEU) to low-enriched uranium (LEU). In order to continue the mission of these reactors, including the Massachusetts of Institute of Technology Reactor (MITR), and achieve similar performance, high-assay low-enriched uranium (HALEU) with a high-density metallic alloy of uranium with 10 wt% molybdenum (U-10Mo) is being evaluated. The impact of the fabrication specification and tolerances was assessed following the preliminary design of the MITR LEU fuel elements using the U-10Mo monolithic alloy. This research focuses on the analysis of fabrication specification impact on thermal hydraulics (TH) characteristic of the MITR LEU core as a function of the variation of the relevant fuel specification parameters (e.g., coolant channel gap thickness, fuel plate thickness, etc.). The analyses are performed based on an all-fresh LEU fuel conversion plan identified in a preliminary safety analysis report submitted to the Nuclear Regulatory Commission. The reactor power margin to the onset of nucleate boiling (ONB) is assessed under the limiting safety system settings (LSSS), where a scram occurs, to ensure there is sufficient margin to the reactor safety limit, which is defined by the onset of flow instability that occurs after the ONB. The best estimate plus uncertainty approach is employed to analyze this TH characteristic, which yields realistic results while maintaining adequate conservatism, utilizing a statistical uncertainty propagation method with the STAT7 code. The TH characteristic is analyzed as a function of the variability of the specification parameters resulting from the fabrication process. The main findings of this study show that the MITR core can meet the TH safety and operational requirements at the all-LEU initial core startup (cycle 1), selected transition cycles (most reactive cycle and most limiting cycle: cycle 3 and 5, respectively) and equilibrium (cycle 14) cores under all limiting fabrication parameter combinations considered. In addition, the analyses show that the dependency of the core power margin to ONB on those specification parameters that have the most direct impact on TH performance is non-linear but monotonically decreasing within the specification tolerances. The third order polynomial fit curves are reported in detail for selected limiting cases and can serve as a powerful tool for future MITR fuel management in cases such as when HALEU supply is established that may allow additional cycle length or other operational benefits.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"219 ","pages":"Article 111376"},"PeriodicalIF":1.9,"publicationDate":"2025-04-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143825795","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 saturated flow boiling of water in vertical helically coiled tubes","authors":"Yuqing Su,&nbsp;Xiaowei Li,&nbsp;Xinxin Wu","doi":"10.1016/j.anucene.2025.111464","DOIUrl":"10.1016/j.anucene.2025.111464","url":null,"abstract":"<div><div>Helically coiled tubes are used in the High Temperature Gas-cooled Reactor (HTGR) Once Through Steam Generator (OTSG). However, due to centrifugal forces and secondary flows, heat transfer characteristics inside these tubes may differ from that in straight tubes. A clear understanding of the heat transfer inside helically coiled tubes is essential for the design and operation of the OTSG. In this study, we experimentally investigated saturated flow boiling in helically coiled tubes with a large curvature ratio (<em>δ</em> = 0.109). The experimental parameters cover a broad range. The system pressure is from 3.5 to 7 MPa, mass flux is from 300 to 1100 kg/(m²·s) and heat flux is from 50 to 600 kW/m². Results show that the inner wall temperature distribution is uneven, with the highest temperature on the inner side and the lowest on the outer side. Increasing heat flux enhances the saturated flow boiling heat transfer coefficient. At low steam quality, the heat transfer coefficient is not significantly affected by mass flux variations. However, at higher steam quality, increasing mass flux improves heat transfer. An increase in system pressure enhances the heat transfer coefficient at lower steam qualities but reduces it at higher steam qualities. Six correlations for the saturated flow boiling were evaluated, with the Gungor-Winterton correlation originally developed for straight tubes showing accurate predictions for heat transfer coefficients in helically coiled tubes (MAPE is 12.72 %, RMS is15.36 %). This indicates that even in helically coiled tubes with a large curvature ratio, no significant difference is observed in the saturated flow boiling heat transfer coefficient compared to straight tubes.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"219 ","pages":"Article 111464"},"PeriodicalIF":1.9,"publicationDate":"2025-04-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143825794","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":"Parameter identification of fluid field based on CFD reduced-order model and 3D-Var data assimilation","authors":"Chuqiao Dai ,&nbsp;Di Yang ,&nbsp;Chunyu Zhang ,&nbsp;Helin Gong","doi":"10.1016/j.anucene.2025.111459","DOIUrl":"10.1016/j.anucene.2025.111459","url":null,"abstract":"<div><div>Data assimilation (DA) significantly improves the accuracy of field state and parameter estimation by merging experimental data with predictions from high-fidelity numerical models. However, despite their precision and high resolution, the substantial computational cost associated with these numerical models often hinders their practical application in DA. To overcome this challenge, this study presents a novel three-dimensional variational (3D-Var) DA framework that leverages a reduced-order model (ROM) for boundary parameter estimation in computational fluid dynamics (CFD) models. The framework utilizes Proper Orthogonal Decomposition (POD)-Galerkin projection to construct the ROM, enabling near real-time solutions and significantly enhancing computational efficiency. Furthermore, a nonlinear observation operator is developed within the reduced basis space, which directly connects parameters with observational data. This approach eliminates the necessity for full-state reconstruction, thereby further streamlining the computational process. Benchmark results indicate that the proposed method achieves high accuracy and robustness, offering optimal background information for subsequent state estimation. This advancement not only reduces computational overhead but also maintains the integrity and reliability of the estimations, making it a promising tool for real-time applications in complex fluid dynamics scenarios.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"219 ","pages":"Article 111459"},"PeriodicalIF":1.9,"publicationDate":"2025-04-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143825796","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 semi-analytical method for modelling station blackout transients in liquid metal-cooled reactors","authors":"Janne Wallenius,&nbsp;Fredrik Dehlin","doi":"10.1016/j.anucene.2025.111414","DOIUrl":"10.1016/j.anucene.2025.111414","url":null,"abstract":"<div><div>A semi-analytical method for modelling station blackout performance in liquid metal reactors is developed, permitting to identify key factors determining peak temperatures during the transient, and hence to design associated passive safety systems. It is shown that integrity of the fuel cladding during this transient can be ensured by adequate dimensioning of coolant channels, the primary system and the vessel air cooling circuit. These dimensions are determined using algebraic equations and postulated values for a minimum/maximum permissible Reynolds number, dimensionless parameters for the fuel cladding tube geometry and heat sink elevation, a guard vessel height, the nominal core power, permitted temperature gradients in the vessel air cooling system and the air cooling system chimney height. The model suggests that the required coolant volume is a rapidly growing function of core power, and that this volume needs to be 40% larger in a sodium-cooled reactor than in a lead-cooled reactor.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"219 ","pages":"Article 111414"},"PeriodicalIF":1.9,"publicationDate":"2025-04-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143825793","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}