Annals of Nuclear Energy最新文献

{"title":"Machine learning-based intelligent models for accurate prediction of key parameters in pressurized water reactors with operational data","authors":"Chenghui Wan ,&nbsp;Wenhui Liu ,&nbsp;Jiahe Bai ,&nbsp;Jianfu Zhang ,&nbsp;Hanbo Gao ,&nbsp;Zhewen Xiao","doi":"10.1016/j.anucene.2025.111518","DOIUrl":"10.1016/j.anucene.2025.111518","url":null,"abstract":"<div><div>In this research, a quick and accurate prediction model has been developed for the key parameters of pressurized water reactors (PWRs) by applying the residual module neural network and single-channel convolutional neural network. The improved Gen-II PWRs, including CNP1000 and CPR1000, constitute the principal operating units in China and have accumulated a great many measurement values for critical boron concentration (CBC) and power distributions. To improve the prediction accuracy of these key parameters, this study utilizes approximately 19,971 sets of measurement values for CBC and 618 sets of data for 3-D power distribution. These datasets are combined with core state parameters provided by the PWR-core analysis code LOCUST/SPARK, which include factors such as fuel-assembly loading, assembly-averaged burnup, and poison-isotope distribution, among others, to train the intelligent models. As a result the intelligent model linking simulated core state parameters to measured key parameters has been established, making it possible to accurately and rapidly predict the measured CBC and power distribution. The verification tests indicate that the predicted CBCs have an average error of about 3.32 ppm with a 95 % confidence interval of [-6.0 ppm, 7.5 ppm], taking only 1.8 ms for each prediction. The predicted axial-power distributions and axial offset (AO) have an average relative error of about 0.46 % and 0.069 %, with 95 % confidence intervals of [-1.4 %, 1.4 %] and [-0.124 %, 0.071 %], respectively, and require only 8.3 ms for each prediction. Additionally, the predicted radial core power distributions have an average error of about 0.31 % with a 95 % confidence interval of [-0.92 %, 0.92 %] and require only 3.2 ms for each prediction. These results demonstrate that the intelligent models established in this research are capable of predicting key parameters with high accuracy and efficiency, which provides a novel technical approach for further enhancing the operational economy and safety of PWR nuclear power plants.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"220 ","pages":"Article 111518"},"PeriodicalIF":1.9,"publicationDate":"2025-05-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143922375","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":"Toward integrating high-fidelity CFD approaches in the thermal-hydraulic analysis of turbulent dry cask systems","authors":"Sinan Okyay ,&nbsp;Elia Merzari ,&nbsp;Derek Gaston ,&nbsp;Fande Kong ,&nbsp;Paolo Balestra","doi":"10.1016/j.anucene.2025.111500","DOIUrl":"10.1016/j.anucene.2025.111500","url":null,"abstract":"<div><div>Nuclear power plants have been supplying resilient and reliable electricity for decades, contributing to energy independence of the U.S.. However, nuclear waste management remains one of the most significant challenges in the industry. The safety of dry cask storage systems relies heavily on their thermal-hydraulic performance. Computational Fluid Dynamics (CFD) simulations are often used to demonstrate this performance and ensure that the system design meets safety standards. This study presents reduced numerical models for various types of dry cask systems. These numerical models can produce efficient and fast results based on the employed modeling strategies. Additionally, the study uses a novel approach to high-fidelity simulations to evaluate modeling assumptions in dry cask modeling. Large Eddy Simulations (LES) are used for this purpose, particularly in regions where fluid velocity is relatively high and the turbulence characteristics become important. The results of these high-fidelity simulations will enhance the interpretation of outcomes produced from a lower-fidelity CFD model.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"220 ","pages":"Article 111500"},"PeriodicalIF":1.9,"publicationDate":"2025-05-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143918386","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":"CVEC: A customized VIS-based equivalence checker for verifying commercial field-programmable gate array synthesis software in small modular reactors","authors":"Yoona Heo ,&nbsp;Sejin Jung ,&nbsp;Eui-Sub Kim ,&nbsp;Junbeom Yoo","doi":"10.1016/j.anucene.2025.111484","DOIUrl":"10.1016/j.anucene.2025.111484","url":null,"abstract":"<div><div>Field-programmable gate array (FPGA) is a hardware-based platform widely used in safety-related systems. FPGA development involves synthesis, placement and routing, with designs ultimately downloaded onto the device. Commercial FPGA synthesis software converts register-transfer level (RTL) designs into gate-level representations. Standards such as NUREG/CR-6421, IEEE Std 7-4.3.2, EPRI NP-5652, and EPRI TR-106439 require verification of these tools, while IEC 62566 mandates static analysis, including equivalence checking, for nuclear power plant applications. This paper introduces <em>CVEC</em>, a customized VIS-based equivalence checker that verifies the correctness of FPGA synthesis software. It performs equivalence checking between RTL designs and gate-level designs synthesized using <em>Synopsys Synplify Pro</em> within the <em>Libero IDE</em>. When verification is successful, it ensures that these softwares operate correctly at the synthesis level. Two case studies demonstrate the effectiveness of <em>CVEC</em> in verifying the functional correctness of commercial FPGA synthesis softwares.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"220 ","pages":"Article 111484"},"PeriodicalIF":1.9,"publicationDate":"2025-05-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143911908","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 fault sensor detection and localization method based on autoencoder","authors":"Longfei Shan,&nbsp;Yongkuo Liu,&nbsp;Xin Ai,&nbsp;Jiarong Gao","doi":"10.1016/j.anucene.2025.111546","DOIUrl":"10.1016/j.anucene.2025.111546","url":null,"abstract":"<div><div>Some sensors in nuclear power plants are exposed to extreme conditions, including high temperatures, high pressures, and intense radiation. As a result, the sensors may experience varying degrees of aging or malfunction. In this paper. we investigate the problem of detecting failures and locating faults in nuclear power plant sensors and proposes a sensor failure detection and localization method based on autoencoders. This method primarily consists of three parts: anomaly detection, fault diagnosis, and faulty sensor localization. Among these, the autoencoder method is used for anomaly detection and faulty sensor localization; feature extraction methods and the random forest algorithm are employed for fault diagnosis. This study used the Fuqing Nuclear Power Plant simulator as the data source to verify the effectiveness of the proposed algorithm. The experimental results show that the autoencoder method can identify faults and normal states with 99.6 % accuracy. The feature extraction-Random Forest method used to distinguish between sensor faults and system faults can identify system faults and sensor faults with 100 % accuracy, and the feature extraction method proposed in this paper is generally applicable, improving accuracy by 2.65 %-14 %. The autoencoder method for fault sensor localization demonstrates good diagnostic accuracy in both single-sensor and multi-sensor fault localization with a low false alarm rate. The autoencoder-based sensor fault detection and localization method demonstrates good diagnostic capability, providing a universal and direct sensor fault diagnosis framework for various types of sensor fault diagnosis.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"220 ","pages":"Article 111546"},"PeriodicalIF":1.9,"publicationDate":"2025-05-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143908271","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":"Simulation of the fuel-assembly bowing in PWR based on the coupling of pin-resolved neutronics and pin-by-pin subchannel","authors":"Runze Liu ,&nbsp;Fan Li ,&nbsp;Zhouyu Liu ,&nbsp;Changyou Zhao ,&nbsp;Xianghui Lu ,&nbsp;Liangzhi Cao ,&nbsp;Hongchun Wu","doi":"10.1016/j.anucene.2025.111520","DOIUrl":"10.1016/j.anucene.2025.111520","url":null,"abstract":"<div><div>In recent years, some nuclear power reactors have experienced power tilts. According to the technical specifications for the operation of the nuclear power plant (NPP), the reactor can only be operated at a power level below 100%, which seriously affects the economic efficiency of NPPs. The bowing of fuel assemblies, which can lead to a power tilt, is investigated in this study, and the simulation method that considers fuel-assembly bowing in high-fidelity neutronics and thermohydraulic coupling is conducted. For the neutronics calculations, the equivalent density method is proposed to simulate the fuel-assembly bowing. The effect of the fuel-assembly bowing on the thermohydraulic feedback was taken into account by changing the flow area of the cooling channels. Some tests were performed to verify the results of the proposed method. The results show that the proposed simulation method can be used to study and analyze the power tilts caused by the fuel-assembly bowing in advanced PWRs.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"220 ","pages":"Article 111520"},"PeriodicalIF":1.9,"publicationDate":"2025-05-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143903580","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":"Investigation of shielding and dosimetric parameters for neutrons emitted from various types of fission sources","authors":"Rahim Khabaz ,&nbsp;Harith Mohamed Al-Azri","doi":"10.1016/j.anucene.2025.111541","DOIUrl":"10.1016/j.anucene.2025.111541","url":null,"abstract":"<div><div>This study evaluates the shielding and dosimetric characteristics of neutrons emitted from various fission sources, utilizing the Monte Carlo simulations. Neutron emissions, crucial in nuclear applications and safety, are analyzed across a spectrum of fissionable isotopes. The work focuses on determining the mean fluence-to-dose conversion coefficients, average total macroscopic cross sections, as well as the flux and dose-equivalent buildup factors for various shielding materials. Results highlight that neutron flux and dose-equivalent buildup factors increase with shield thickness, influenced significantly by the shielding material’s effective mass number. Notably, concrete exhibits the highest buildup factor, contrasting with water, which shows the lowest. Furthermore, polynomial equations parameterized by source and material coefficients facilitate accurate flux and dose-equivalent buildup factor calculations. This study enhances the understanding of neutron shielding effectiveness and dosimetry, which is crucial for optimizing radiation protection measures in various nuclear contexts, from reactor operations to medical applications.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"220 ","pages":"Article 111541"},"PeriodicalIF":1.9,"publicationDate":"2025-05-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143903658","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":"Effect of permutation and combination patterns on the performances of shielding materials","authors":"Dengjian Wu ,&nbsp;Junjun Gong ,&nbsp;Zifu Hao ,&nbsp;Chengqiang Liang","doi":"10.1016/j.anucene.2025.111465","DOIUrl":"10.1016/j.anucene.2025.111465","url":null,"abstract":"<div><div>Composite materials have been extensively employed to enhance the performance of shielding materials under the condition of neutron and γ radiation fields. Researchers generally design the component based on the energy spectrum information of the radiation field with Monte Carlo software, then achieve the uniform mixing of components with specific technological processes, ultimately obtaining the composite shielding material with expected performance. However, due to the substantial differences in properties between the components of composite materials, the process of mixing them uniformly is complex and costly. This drawback may be unacceptable in specific situations such as scenarios requiring extensive use and complex on-site installation environments. Lead boron polyethylene (PbBPE), a widely studied and utilized composite material, was analyzed using the Geant4 software in conjunction with the K-Nearest Neighbors (KNN) algorithm to calculate and analyze the performance of laminated materials with the same mass fraction and total mass but with different permutations. The results indicate that the performance of the sandwich-like laminated material formed by filling boron polyethylene (BPE) between two layers of lead (Pb) plates may be superior to that of the composite material PbBPE.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"220 ","pages":"Article 111465"},"PeriodicalIF":1.9,"publicationDate":"2025-05-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143903659","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":"Influence of level density models on proton and deuteron-induced reactions using zinc target for the production of 66-68Ga medical radioisotopes","authors":"Bassam T. Al-Azraq ,&nbsp;Raghad I. Mahmood ,&nbsp;Lubna Abduljabbar Mahmood ,&nbsp;Radhwan Ch. Mohsin ,&nbsp;Rusul S. Jaffer","doi":"10.1016/j.anucene.2025.111535","DOIUrl":"10.1016/j.anucene.2025.111535","url":null,"abstract":"<div><div>Due to their medical importance in imaging and therapy, this paper examines proton- and deuteron-induced reactions on zinc targets from a theoretical perspective. Medically important ⁶⁶Ga, ⁶⁷Ga, and ⁶⁸Ga were simulated using TALYS 2.0 and different level density models. The relative variance technique was used to assess model agreement with experimental data given in EXFOR library. Key production parameters, including radionuclidic impurities, optimal energy range, theoretical yield, and target thickness, were derived from the selected theoretical model, which was chosen based on its agreement with experimental trends. Results confirmed the models’ predictive accuracy across a wide energy range in estimating the nuclear reaction cross-sections for ^66,67,68Ga production. For a number of studied reactions, the TGHFB model specifically agreed very well with the experimental data. Zinc targets proved effective and feasible for producing ^66,67,68Ga at low proton and deuteron energies, as typically available in medical cyclotrons.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"220 ","pages":"Article 111535"},"PeriodicalIF":1.9,"publicationDate":"2025-05-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143903660","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":"Lead-cooled fast reactor SGTR accident pressure wave analysis and structural response","authors":"Xi Huang ,&nbsp;Chong Qin ,&nbsp;Kefan Zhang ,&nbsp;Lixiang Zhang ,&nbsp;Hongli Chen","doi":"10.1016/j.anucene.2025.111489","DOIUrl":"10.1016/j.anucene.2025.111489","url":null,"abstract":"<div><div>Steam Generator Tube Rupture (SGTR) accidents are one of the most serious types of accidents in Lead-cooled Fast Reactors (LFRs). Due to the large pressure and temperature differences between the two sides of the steam generator heat transfer tubes in lead–bismuth fast reactors, combined with the corrosive effects of liquid lead–bismuth eutectic (LBE), steam generator heat transfer tube rupture accidents may occur. In this paper, the SGTR accident in the lead-cooled fast reactor M²LFR-1000 is taken as an example. The computational fluid dynamics (CFD) program is used to simulate the early stage of SGTR accident, and the correctness of the model is verified through experiments. The severity of the accident under different rupture sizes is also assessed. By coupling ANSYS Fluent with ANSYS Structural, the impact of a single heat exchanger tube rupture on neighboring tubes is calculated to evaluate the possibility of chain rupture accidents. The results of the study show that as the pressure wave peak increases with larger rupture sizes, the risk of localized fracture is higher in internally threaded tubes than in smooth tubes, as determined by stress assessment.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"220 ","pages":"Article 111489"},"PeriodicalIF":1.9,"publicationDate":"2025-05-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143903581","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":"Nuclear hydrogen demonstration project using the HTTR – Demarcation of nuclear-industrial laws and design standards","authors":"Takeshi Aoki,&nbsp;Atsushi Shimizu,&nbsp;Katsunori Ishii,&nbsp;Keisuke Morita,&nbsp;Naoki Mizuta,&nbsp;Kaoru Kurahayashi,&nbsp;Takanori Yasuda,&nbsp;Hiroki Noguchi,&nbsp;Yasunobu Nomoto,&nbsp;Kazuhiko Iigaki,&nbsp;Hiroyuki Sato,&nbsp;Nariaki Sakaba","doi":"10.1016/j.anucene.2025.111503","DOIUrl":"10.1016/j.anucene.2025.111503","url":null,"abstract":"<div><div>A high temperature gas cooled reactor has inherent safety features and potential to produce competitive and large amounts of carbon-free hydrogen. Japan Atomic Energy Agency has initiated the high temperature engineering test reactor (HTTR) Heat Application Test Project to establish coupling technologies for a nuclear hydrogen production system that connects a nuclear reactor plant and a hydrogen production plant. In order to ensure the public safety, appropriate laws and design standards should be applied to the hydrogen production plant because it handles combustible and toxic gases. The present study proposed a relative evaluation methodology using six metrics to characterize the superiority of candidates for the demarcation of applicable laws and design standards for a for the nuclear hydrogen production system. The proposed methodology was applied to the HTTR Heat Application Test Facility to select the most superior demarcation of applicable laws and design standards from six candidates for the HTTR Heat Application Test Facility. Candidates, applying nuclear law to all facilities, showed least superiority due to the higher cost of commercialized hydrogen production and the difficulty in entry for non-nuclear vendors to the business. The candidates, applying the High Pressure Gas Safety Act for the steam reformer and the Heat Application Test Facility (hydrogen production plant), showed least superiority in the feasibility of the system or the demonstration of the key equipment and technologies for commercialization depending on the installation of the secondary intermediate heat exchanger. On the other hand, a candidate applying the High Pressure Gas Safety Act to the Heat Application Test Facility (hydrogen production plant) and design standards established under the High Pressure Gas Safety Act to the steam reformer did not show the lowest category in any of the metrics, and was proposed as the most superior candidate for the demarcation of applicable laws and design standards the HTTR Heat Application Test Facility.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"220 ","pages":"Article 111503"},"PeriodicalIF":1.9,"publicationDate":"2025-05-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143900302","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}