Annals of Nuclear Energy最新文献

{"title":"Actor-Critic deep reinforcement learning-based integrated fault diagnosis method for nuclear power plants","authors":"Tong Li ,&nbsp;Sichao Tan ,&nbsp;Jiangkuan Li ,&nbsp;Ruifeng Tian ,&nbsp;Jihong Shen ,&nbsp;Jiaoshen Xu","doi":"10.1016/j.anucene.2025.111907","DOIUrl":"10.1016/j.anucene.2025.111907","url":null,"abstract":"<div><div>Although the international community has carried out a lot of research work on fault diagnosis, the nonlinear and high coupling characteristics of nuclear reactors and complex transient conditions put forward higher requirements for fault diagnosis algorithms. In this paper, an integrated method based on reinforcement learning is proposed, in which the base models of the integrated model are pre-trained machine learning units, and Twin delayed deep deterministic policy gradient is used to learn dynamic weight update strategies to adjust the diagnosis probability of the base models based on One-vs-Rest method. An idea of enhanced value propagation is designed and applied to the reinforcement learning. Experimental results show that the fault diagnosis integrated framework based on reinforcement learning has higher accuracy and lower misdiagnosis rate, the enhanced value propagation is proved to be able to obtain higher fault diagnosis accuracy with faster convergence and stronger optimization ability.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"227 ","pages":"Article 111907"},"PeriodicalIF":2.3,"publicationDate":"2025-10-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145236288","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":"Modeling and multifactor coupling sensitivity analysis of butterfly check valve under the station blackout of PWR","authors":"Fulong Tang ,&nbsp;Wei Wang ,&nbsp;Yiming Luo","doi":"10.1016/j.anucene.2025.111906","DOIUrl":"10.1016/j.anucene.2025.111906","url":null,"abstract":"<div><div>A butterfly check valve is an important part in the safe operation of the pressurized water reactor(PWR). To study the influence of the butterfly valve on coolant flow characteristics under multi-factor coupling and evaluate the post-accident processes of the reactor, a physical model of the butterfly valve is established based on MELCOR 1.8.5 and validated by experiments. The equivalent valve opening share model and the equivalent resistance coefficient model are established to simplify the modeling process. Through sensitivity analysis, the effects of different modeling methods on flow rate and accident process are evaluated. Additionally, the butterfly check valve model is applied to analyze the transient response of the PWR during the station blackout(SBO) accident. The results show that the equivalent valve opening share model has higher accuracy in the sensitivity analysis. The butterfly valve model under multi-factor coupling enables accurate simulation of post-accident scenarios while significantly enhancing calculation accuracy.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"227 ","pages":"Article 111906"},"PeriodicalIF":2.3,"publicationDate":"2025-10-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145236287","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":"Stochastic, multi-path vulnerability assessment of a physical protection system using non-fixed critical detection points","authors":"Melih Ozkutuk ,&nbsp;Sunil S. Chirayath","doi":"10.1016/j.anucene.2025.111914","DOIUrl":"10.1016/j.anucene.2025.111914","url":null,"abstract":"<div><div>Physical protection system (PPS) at nuclear facilities must be assessed against diverse adversary strategies, uncertainties in detection performance, and potential insider actions. Traditional estimate of adversary sequence interruption (EASI) model assumes fixed critical detection points (CDPs) and fail to capture detection variability or multi-path vulnerabilities. This paper introduces a stochastic, multi-path framework with non-fixed CDPs (nf-CDPs) that accounts for uncertainty in detection probability, communication reliability, and response delays. A stochastic approach (100,000 simulations) is applied to adversary path generation under five adversary strategies: random, rushing, covert, deep penetration, and most vulnerable path (MVP). The framework incorporates simplified insider modeling and cost–performance analysis. Results show nf-CDPs shift dynamically with stochastic sampling, producing wider probability of interruption (P_I) distributions than fixed-point CDP assumptions. Sensitivity analysis highlights insider presence and response force variability, while regression confirms a nonlinear cost–P_I relationship. The study demonstrates nf-CDPs provide a more realistic PPS assessment and practical recommendations.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"227 ","pages":"Article 111914"},"PeriodicalIF":2.3,"publicationDate":"2025-10-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145236286","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":"Applying the derivative source method to flux derivatives in Monte Carlo fixed source problems caused by interface shifts between spontaneous fission materials","authors":"Toshihiro Yamamoto ,&nbsp;Hiroki Sakamoto","doi":"10.1016/j.anucene.2025.111916","DOIUrl":"10.1016/j.anucene.2025.111916","url":null,"abstract":"<div><div>Derivative source method (DSM) determines neutron flux derivatives by solving the transport equation obtained by differentiating the fixed source Boltzmann neutron transport equation with respect to cross sections or interface positions. This study proposes a method for analyzing neutron flux derivatives using the DSM when the interface between spontaneous fission materials shifts. Since the cross section and source distribution in this case are given by Heaviside functions, the derivative due to the interface shift is obtained using a Dirac delta function at the interface. In the DSM, derivative source particles given by the delta function are emitted from both sides of the interface and undergo a random walk; however, we propose a method, applicable to multi-group Monte Carlo calculations, to integrate these two into a single particle. The DSM is conducted in the course of an ordinary fixed source calculation for cross section changes due to interface shifts. During the calculation, each time a particle crosses a shifting interface, a derivative source particle is emitted from the crossing point and its random walk is subsequently performed. Furthermore, an additional calculation is conducted when the interface of spontaneous fission source material shifts, in which derivative source particles are emitted from the interface due to the spontaneous fission source shift and subjected to a random walk. Test calculations demonstrate the DSM’s effectiveness in terms of computational accuracy and efficiency in comparison to neutron flux derivatives obtained from differences in neutron flux before and after a minute displacement of the interface.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"227 ","pages":"Article 111916"},"PeriodicalIF":2.3,"publicationDate":"2025-10-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145236289","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":"Primary frequency control of multi-module high temperature gas-cooled reactor cogeneration unit based on active disturbance rejection control","authors":"Congcong Li,&nbsp;Zhe Dong,&nbsp;Weidong Sun,&nbsp;Xiaojin Huang","doi":"10.1016/j.anucene.2025.111816","DOIUrl":"10.1016/j.anucene.2025.111816","url":null,"abstract":"<div><div>Multi-module high temperature gas-cooled reactor (mHTGR) has gained engineering significance when the world’s first demonstration project, HTR-PM, went commercial in 2023, and multiple aspects of application for mHTGR has been designed and verified in recent years. Certain nuclear power plant cogeneration unit based on mHTGR can be utilized to generate hydrogen or unirradiated steam to industrial consumer for various purposes through the utilizing of the high steam temperature. The concept of coordinated control is raised during the design and verification process, and prospers when the collaboration between mHTGR and various industrial processes enlarges. Control algorithm is one of the essentials of maintaining safe and stable operation of a nuclear power plant in a big picture. Adaption from mature control theory proves to attain better performance and stronger robustness other than the conventional control method. Active disturbance rejection control (ADRC) is one of the mature theories that has been utilized as a countermeasure against internal or external disturbance of various types, and can actively compensate the disturbance using extended state observer. In this paper, a nonlinear ADRC is proposed and implemented as primary frequency control (PFC) in certain mHTGR cogeneration unit, and results are compared via simulations and verification of the control performance is validated when subjected to disturbances at various locations within the model of the mHTGR cogeneration unit. Through comparisons, ADRC proves to improve the performance and robustness of PFC as expected, and requires no further information from the mHTGR cogeneration unit, easing the implementation of ADRC.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"226 ","pages":"Article 111816"},"PeriodicalIF":2.3,"publicationDate":"2025-10-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145216783","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":"Data-fusion-based variable-fidelity reduced-order model for accurate thermal–hydraulic behavior prediction in a 6 × 6 rod bundle","authors":"Guangyun Min ,&nbsp;Xiuzhong Shen ,&nbsp;Jingtao Xue ,&nbsp;Laishun Wang ,&nbsp;Naibin Jiang","doi":"10.1016/j.anucene.2025.111905","DOIUrl":"10.1016/j.anucene.2025.111905","url":null,"abstract":"<div><div>Achieving fast and low-cost computation of the thermal–hydraulic flow field inside a nuclear reactor is of great significance for understanding the reactor’s thermal–hydraulic characteristics and ensuring its operational safety. In this study, a variable-fidelity reduced-order model (ROM) is proposed, which integrates data from both low-fidelity and high-fidelity simulations. The low-fidelity model—characterized by a coarse mesh, standard <em>k</em>-<em>ε</em> turbulence model, Semi-Implicit Method for Pressure Linked Equations (SIMPLE) algorithm, and first-order upwind scheme—is employed to capture the global trend of the flow field of a 6 × 6 rod bundle. Due to its low computational cost, a large number of low-fidelity samples can be generated efficiently. In contrast, the high-fidelity model—using a refined mesh, SST <em>k</em>-<em>ω</em> turbulence model, coupled solver, and second-order upwind scheme—is utilized to correct the flow fields obtained from the low-fidelity model. However, the high computational cost limits the number of high-fidelity samples. To bridge the fidelity gap, a bridge function is constructed to fuse the data from both fidelities. Proper Orthogonal Decomposition (POD) is applied to extract POD modes and corresponding POD coefficients from both low-fidelity and high-fidelity snapshots. The difference between the POD coefficients of the two fidelities is modeled as a function of the input design variables using a Radial Basis Function (RBF) surrogate. For any new design variables, the differences in POD coefficients can be rapidly predicted and added to the corresponding low-fidelity POD coefficients. These corrected POD coefficients are then combined with the high-fidelity POD modes to predict the flow fields. Compared to conventional ROMs, the proposed data-fusion-based variable-fidelity model offers improved computational efficiency and prediction accuracy. The results of this study contribute to the efficient and accurate thermal–hydraulic analysis and optimization design of nuclear reactors.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"226 ","pages":"Article 111905"},"PeriodicalIF":2.3,"publicationDate":"2025-10-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145216781","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":"Transient level determination with machine learning for pressurized water reactor VVER-1000","authors":"Ceyhun Yavuz,&nbsp;Senem Şentürk Lüle","doi":"10.1016/j.anucene.2025.111908","DOIUrl":"10.1016/j.anucene.2025.111908","url":null,"abstract":"<div><div>Nuclear reactors carry significant risks of severe consequences in the event of accidents. Therefore, the imperative of ensuring their safe operation is paramount. During transient conditions, key parameter time series may fluctuate in ways that are not immediately visible. Therefore, the detection of these transients is essential for preventive measures and protective actions. This work focuses on identification of 53 transient sub-scenarios derived from reactivity insertion via rod withdrawal, steam leak from pressurizer, loss of flow and loss of coolant accidents affecting both hot and cold legs main transient for VVER type reactor. 91 features have been handled for model assessment with 465,465 data points. K-Nearest Neighbor, Decision Tree Classifier, Random Forest Classifier, Gradient Boosting, Logistic Regression, Support Vector Machine, Naïve Bayes and Multilayer Perceptron methods were applied for three different approaches. A one-step, two-steps, and grouped one-step approaches were considered for identification sub-scenarios. In the one-step approach, the exact sub-scenario (e.g. 25% rod withdrawal) was identified. In the two steps approach, the main transients (e.g. rod withdrawal) were identified first and then the sub-scenario (e.g. 25% rod withdrawal) were identified. In the grouped one-step approach, sub-scenarios were grouped (e.g. 20 to 30% rod withdrawal) to increase the accuracy of predictions. While the highest accuracy in one-step approach was 74.66%, two-steps approach had 99.51% main transient identification but 86.44% total accuracy. The grouping of sub-scenarios achieved a less precise but more accurate result with 92.33% accuracy. In conclusion, fast transient identification for VVER type reactors was achieved with two-steps approach.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"226 ","pages":"Article 111908"},"PeriodicalIF":2.3,"publicationDate":"2025-10-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145216782","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":"Energy management systems in microgrids and future prospects application in nuclear power plants: a review","authors":"Alaa B. Maraey ,&nbsp;Mohamed F. Kotb ,&nbsp;Abdelhady Ghanem ,&nbsp;Mohamed Elgohary","doi":"10.1016/j.anucene.2025.111904","DOIUrl":"10.1016/j.anucene.2025.111904","url":null,"abstract":"<div><div>Nowadays, microgrids (MGs) become one of the most recent interest of researchers and investors as a sustainable energy system. As MG is a group of distributed energy resources (DERs) with different intermittent properties, big challenges are presented when either integrating with conventional resources or renewable energy to feed different types of loads within a specified area. To face these encounters, MGs and energy management system (EMS) installations, operation, control performance, quality, resilience and security for such complicated system should be understood, analyzed and satisfied. This article presents a comprehensive review on the structure of the MGs, planning strategies, close monitoring systems along with the utilized interactive control approaches and optimization strategies that were developed for the MGs’ optimal operation. Also, the classifications of EMS based on supervisory control, and the approach for making a decision are presented along with their advantages and disadvantages. The role of MGs to introduce a reliable power supply for loads in different applications, especially critical loads are introduced. Among these applications is the emergency power supply for the critical and safety loads inside nuclear power plants (NPP). The utilization of EMS strategies and decisions in supporting NPP is discussed, to ensure they do not violate its safety rules. The integration between EMS, the NPP standards and the regulatory guides issued by the International Atomic Energy Agency and other international entities are presented. To attain the goals mentioned, more than 260 articles are handled but only 172 papers are introduced in this review. This article may be used as a guide for those interested in power system planning, control, improving power system operation and reliability and utilization of the MGs inside NPPs.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"226 ","pages":"Article 111904"},"PeriodicalIF":2.3,"publicationDate":"2025-10-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145216691","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 1D-CFD coupling method for natural circulation analyses through benchmark analyses of shutdown heat removal tests in EBR-II","authors":"Kazuo Yoshimura ,&nbsp;Norihiro Doda ,&nbsp;Masaaki Tanaka ,&nbsp;Tatsuya Fujisaki ,&nbsp;Satoshi Murakami","doi":"10.1016/j.anucene.2025.111896","DOIUrl":"10.1016/j.anucene.2025.111896","url":null,"abstract":"<div><div>At the Japan Atomic Energy Agency, a multilevel simulation (MLS) system, which enables consistent evaluation from whole plant behavior to local phenomena in the plant components, is being developed to attempt plant design and enhance the safety of sodium-cooled fast reactors. Whole plant and local multidimensional thermal–hydraulic behaviors were evaluated by coupling the in-house one-dimensional plant dynamics analysis code named Super-COPD (1D) and the computational fluid dynamics (CFD) code of ANSYS Fluent. Both codes were coupled and controlled using a Python script-based program. In this study, numerical analyses of the protected and unprotected loss-of-flow tests: SHRT-17 and SHRT-45R, conducted in EBR-II, were performed to validate the coupling method in the MLS system. In the analyses, the cold pool, upper plenum, and Z-shaped pipe connecting the upper plenum and intermediate heat exchanger were modeled by the CFD code. The flow network model for the 1D contained components in the primary heat transport system. By comparing the results of the 1D-CFD coupled analyses with those of standalone analyses using the 1D code and measured data, the validity of the 1D-CFD coupling method for plant dynamics behavior was confirmed. Through numerical analyses, thermal stratification, which is difficult to evaluate using only the 1D code, was clarified in the region modeled by the CFD code. Furthermore, the temperature profiles along the thermocouple trees installed in the upper plenum and cold pool were almost reproduced.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"226 ","pages":"Article 111896"},"PeriodicalIF":2.3,"publicationDate":"2025-09-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145216690","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 assessment of the Mizton multipurpose nuclear microreactor: Design alternatives and performance","authors":"Karina Cruz-Vázquez,&nbsp;Emiliano Morones-García,&nbsp;Juan-Luis François","doi":"10.1016/j.anucene.2025.111903","DOIUrl":"10.1016/j.anucene.2025.111903","url":null,"abstract":"<div><div>Mizton is a multipurpose nuclear microreactor developed by a research group of the School of Engineering of the National Autonomous University of Mexico, currently in the conceptual design stage. The reference microreactor design has a thermal power of 15 MW, using heat pipes with sodium as a coolant, 19.75 % ²³⁵U enriched uranium nitride (UN) TRISO particles as fuel, a monolith of SiC and a secondary reflector of Zr₃Si₂. This research proposes an alternative design with a monolith of graphite, a secondary reflector of ZrC, and a fuel of uranium–plutonium nitride; four models were analyzed in total. The Monte Carlo code Serpent, version 2.1.32 and the JEFF-3.1 cross-section library were used for neutronic simulations. For each of the models, the behavior of the effective neutron multiplication factor and the effect on the reactivity of the variation of the density of the monolith were analyzed. Furthermore, the primary safety parameters such as the Doppler coefficient, the control rods’ worth, the delayed neutron fraction and the neutron generation time were also calculated. In addition, the fuel evolution over a given period at full power was analyzed for each of the models studied. According to the results, the alternative design achieved higher effective neutron multiplication factor values than the reference design. For all the models, the control rods inserted enough reactivity for the safe shutdown, the Doppler coefficient was negative, and the effect on the reactivity of the variation of the monolith density was negligible. The alternative design with enriched UN fuel achieved a longer operating cycle of approximately 9.9 years and reached a burnup of 19,224 MWd/tU.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"226 ","pages":"Article 111903"},"PeriodicalIF":2.3,"publicationDate":"2025-09-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145216689","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}