Annals of Nuclear Energy最新文献

{"title":"Experimental research on fluid elastic instability in tube bundles subjected to single-phase water cross-flow","authors":"Yibo Yin ,&nbsp;Junying Hong ,&nbsp;Puzhen Gao ,&nbsp;Hui Cheng ,&nbsp;Junshuai Sun ,&nbsp;Jiming Wen ,&nbsp;Ruifeng Tian","doi":"10.1016/j.anucene.2025.111501","DOIUrl":"10.1016/j.anucene.2025.111501","url":null,"abstract":"<div><div>The majority of failures in steam generators in the nuclear power plant can be attributed to flow-induced vibrations (FIV) resulting from the cross-flow on the shell side. Understanding the FIV characteristics of the tube bundle plays a significant key in engineering issues, such as the design of tube bundle arrangements and safety verification of equipment. This paper conducts a detailed investigation into the FIV characteristics of three types of arrays with normal triangular, parallel triangular, and rotated non-regular triangle in the cross flow. The effects of the array configuration, pitch ratio, and stiffness of surrounding tubes on FIV were analyzed. The theoretical model was modified based on the experimental data. The reduced velocity (<em>V</em>_p) was increased in a stepped manner during the experiment until the tube bundle was damaged, which led to the termination of the experiment. The range of <em>V</em>_p is 0–3.5. This study analyzed the frequency-domain and time-domain characteristics of three stages: turbulence excitation, vortex shedding, and fluid-elastic instability (FEI). The results indicate that within the range of 1.6 to 2.5, as the pitch ratio increases, the allowable inlet flow velocity becomes larger, and the tube bundle becomes more stable. Under small pitch ratios (less than 1.6), FEI is more likely to occur in the central tube. When the pitch ratio is larger (greater than 2.5), the FEI occurs simultaneously in the central tube and the downstream tube. It is also shown that there is a weak correlation between the <em>V</em>_pcr of transverse FEI and the pitch ratio. There is a strong correlation between the <em>V</em>_pcr of streamwise FEI and the pitch ratio. The coupling between tubes has a significant impact on the occurrence of streamwise FEI. Arranging rigid tubes around the tube is more stable than arranging flexible tubes. Based on the experimental data, the new theoretical model considers the specific variation of the <em>V</em>_pcr with the pitch ratio for rotating non-regular triangular structures. The instability constant is 2.54, and the exponent constant is 0.5.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"219 ","pages":"Article 111501"},"PeriodicalIF":1.9,"publicationDate":"2025-04-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143860725","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":"Improves the stability of containment leakage rate in constant pressure method by linear fitting analysis model","authors":"Li Jianfa ,&nbsp;Chen Guangheng ,&nbsp;Cong Jidong ,&nbsp;Lou Taishan ,&nbsp;Meng Zhaoming","doi":"10.1016/j.anucene.2025.111499","DOIUrl":"10.1016/j.anucene.2025.111499","url":null,"abstract":"<div><div>The constant pressure method for measuring the leakage rate of containment in nuclear power plants is of great significance in high leakage and low-pressure plant environments for its simple testing process, short testing period and high reliable results Therefore, conducting research on the constant pressure method for measuring the leakage rate of containment in large spaces is of great significance. In order to explore a more stable constant pressure method for analyzing the leakage rate of containment, this study develops a theoretical model for calculating the leakage rate of containment by using the Linear fitting analysis model. In the study, many different volumes of containment vessel simulation bodies are used for experimental verification and application research, and the leakage rate data calculated by linear fitting method and mean analysis method are compared. Results show that the leakage rate fluctuation ratio of the Linear fitting method decreases by 51%, and the uncertainty decreases by 87%. Research has confirmed that the Linear fitting method has good stability, high reliability, and low sensitivity to the environment. Based on above advantages, the Linear line fitting analysis model proposed in this study can further support the engineering application of the constant pressure method for measuring the leakage rate of containment.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"219 ","pages":"Article 111499"},"PeriodicalIF":1.9,"publicationDate":"2025-04-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143852206","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":"PINN and KAN temperature prediction of carbon Fiber/Epoxy composite materials irradiated by nuclear environment simulated intense heat fluxes","authors":"Xiaoxiang Han ,&nbsp;Jun Li ,&nbsp;Lin Yuan ,&nbsp;Xin Zhang ,&nbsp;Weijie Zhu ,&nbsp;Yang Liu ,&nbsp;Haiyang Zhang ,&nbsp;Boyu Wang","doi":"10.1016/j.anucene.2025.111454","DOIUrl":"10.1016/j.anucene.2025.111454","url":null,"abstract":"<div><div>The carbon fiber/epoxy composite (CFEC) materials in nuclear environments must endure extreme temperature conditions, making the understanding of their thermal response characteristics crucial for material design and safety assessment. Here, the physical information neural network (PINN) was used to develop a fast solver, while the Kolmogorov Arnold network (KAN) was employed to analyze and predict the correlation between multiple experimental parameters and the material’s surface temperature. An ensemble learning method was adopted to integrate six machine learning models and enhance the robustness of prediction. The surface temperature characteristics of CFEC are significantly influenced by radiation and cooling processes in high- temperature environments. The KAN established an explicit functional relationship between <em>T</em>_max, <em>t</em>_max and input parameters, providing a theoretical basis for optimizing material design. This research contributes to the developing of standards for CFEC performance evaluation under intense heat fluxes, ensuring product quality and safety.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"219 ","pages":"Article 111454"},"PeriodicalIF":1.9,"publicationDate":"2025-04-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143890471","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":"Preparation of UCO sols suitable for gelation into microspheres by sol–gel process with glucose","authors":"Wenjing Zhang ,&nbsp;Chao Yan ,&nbsp;Xiaohe Yu ,&nbsp;Changqing Cao ,&nbsp;Jun Lin ,&nbsp;Guobao Zhang ,&nbsp;Bolong Guo ,&nbsp;Fan Zhang ,&nbsp;Xiuyan Dong ,&nbsp;He Huang","doi":"10.1016/j.anucene.2025.111494","DOIUrl":"10.1016/j.anucene.2025.111494","url":null,"abstract":"<div><div>The design, development, and fabrication of fuel elements characterized by their safety, reliability, extended service life, and high burnup capabilities constitute a critical component within the scope of fourth-generation reactor research and development endeavors, the uranium carbon oxide (UCO) fuel stands out with its superior thermal conductivity and a higher uranium content compared to the more traditional uranium dioxide, making it an ideal candidate for these advanced nuclear reactors. The fabrication of such uranium-based fuel kernels hinges on a crucial step, which is the preparation of the uranium sol, which significantly affects the dispersion of the gel particles and, ultimately, the quality of the final products.</div><div>The objective of this study was to prepare UCO ceramic microspheres using glucose as a carbon source by the external gelation method combined with carbothermal reduction. Additionally, the results demonstrated that uranium sol could be dispersed into gel spheres at a reaction temperature of 80 °C, the molar ratio of urea to uranium (CO(NH₂)₂/U(VI)) is between 3.1–3.3, and the uranium concentration of 1.6–2.4 M. Moreover, adding an appropriate amount of tetrahydrofurfuryl alcohol was proven effective in resolving the cracking issue observed in the gel spheres. Finally, UCO ceramic microspheres with homogeneous microstructures could be prepared through appropriate dispersion and heat treatment.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"219 ","pages":"Article 111494"},"PeriodicalIF":1.9,"publicationDate":"2025-04-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143850271","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Development and application of a neutronics/thermal-hydraulics coupling code based on system code and Anderson acceleration","authors":"Bowen Yang,&nbsp;Jianqiang Shan,&nbsp;Li Ge","doi":"10.1016/j.anucene.2025.111490","DOIUrl":"10.1016/j.anucene.2025.111490","url":null,"abstract":"<div><div>In a main steam line break (MSLB) accident, the degree of mixing of coolant with different temperatures in the lower plenum can significantly affect the power distribution of core, which in turn impacts core safety. The point kinetics model used in traditional system codes is unable to simulate spatial variations in power. To investigate this phenomenon in greater detail, it is essential to couple the system code with a 3D neutron dynamics code. In the development of neutronics/thermal-hydraulics coupling codes based on system codes, operator splitting (OS) method and Picard iteration are often used to advance the time-stepping. The operator splitting method has lower accuracy, while the Picard iteration is an implicit coupling method to improve the computational accuracy of neutronic and thermal–hydraulic coupled codes. However, the convergence of Picard iteration is usually poor, the use of relaxation factors is required to ensure numerical stability. The selection of relaxation factors is typically a manual process, and the optimal range varies under different operating conditions. To enhance the convergence and efficiency of alternate calculations between the neutronic field and thermal–hydraulic field, Anderson acceleration was introduced into the coupled transient simulation of the system code NUSOL-SYS. The developed code with Anderson acceleration was used in transient numerical simulations, and the results demonstrated a significant reduction in the number of coupling iterations. For time steps involving rapid transient changes, the iteration count was reduced to approximately 76.5% of the original Picard iteration, while also eliminating the need for manual selection of relaxation factors. Further validation was performed using the NEACRP 3-D LWR core transient benchmark, where the computed results were in close agreement with reference solutions. Subsequently, the coupled code was applied to simulate the MSLB accident in a pressurized water reactor (PWR). The impact of coolant mixing in the lower plenum on transient parameters was investigated. The results revealed that increased coolant mixing led to lower return to power and reduced local power peaks.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"219 ","pages":"Article 111490"},"PeriodicalIF":1.9,"publicationDate":"2025-04-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143847824","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 design and calibration of a new differential calorimetric dedicated to nuclear heating measurements","authors":"Junxin Zhang ,&nbsp;Tianchi Ma ,&nbsp;Liang Zhang ,&nbsp;Xingjie Peng ,&nbsp;Guofeng Qu ,&nbsp;Shuai Jin ,&nbsp;Yixiong Sun ,&nbsp;Jifeng Han ,&nbsp;Guang Zhao","doi":"10.1016/j.anucene.2025.111485","DOIUrl":"10.1016/j.anucene.2025.111485","url":null,"abstract":"<div><div>Nuclear heating rates are key parameters for the design of materials irradiation experiments. A novel mobile differential calorimeter (CAMORE) has been designed and tested for the High Flux Engineering Test Reactor (HFETR). Monte Carlo simulations and 3D Computational Fluid Dynamics are used for structural and thermal design, and stainless steel was chosen as the sample material to have best precision. An inner electrical heating assembly up to 30 W is set inside the CAMORE to simulate the nuclear heating. The performance of the manufactured calorimeter have been tested in lab, the thermal response sensitivity of 35.57°C·g·W^–1 (@ 3 W·g⁻¹) have been achieved, and adhering to safety temperature constraints. The calibration curves are obtained using the electrical heating, the difference between the sample and reference cell is less than 2 %, which are well agreed with simulations, confirming the suitability for measuring nuclear heating rates within the reactor.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"219 ","pages":"Article 111485"},"PeriodicalIF":1.9,"publicationDate":"2025-04-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143847751","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 simulating of the gas flow inside a molecular pump with different geometries by Monte Carlo method to predict the compression ratio","authors":"Masoud Khajenoori ,&nbsp;Jaber Safdari ,&nbsp;Sadegh Yousefi-Nasab","doi":"10.1016/j.anucene.2025.111391","DOIUrl":"10.1016/j.anucene.2025.111391","url":null,"abstract":"<div><div>The molecular pump is a fixed piece in the outer space of the rotor in a gas centrifuge. The molecular pump has grooves that, when moving particles collide with it, regarding the grooves are diverted in the direction that will be routed back into the rotor. One of the molecular methods for analyzing the flow of gas inside the molecular pump’s groove is the DSMC method. In this paper, while modifying the Sickafus analytical method, a trapezoidal groove of a molecular pump and a molecular pump with full geometry have been simulated using the dsmcFoam solver in OpenFOAM. The results of DSMC simulation have been compared with the results of the modified Sickafus analytical method. Also, simulation of the molecular pump with rectangular, triangular, and semi-circular geometries has been performed, and the results have been compared with the analytical method. The results show that the semicircular geometry creates the highest compression ratio during the molecular pump. Also, the results show that by improving the Sickafus analytical method, it is possible to achieve an accurate answer with a high-speed analytical method without the need to simulate the molecular pump with complex and high computing time method such as DSMC.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"219 ","pages":"Article 111391"},"PeriodicalIF":1.9,"publicationDate":"2025-04-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143844186","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Development and application of three-dimensional multi-physics and multi-scale coupling program for lead cooled fast reactor","authors":"Sifan Dong,&nbsp;Jingguo Wei,&nbsp;Weixiang Wang,&nbsp;Kefan Zhang,&nbsp;Rui Pan,&nbsp;Shuai Wang,&nbsp;Hongli Chen","doi":"10.1016/j.anucene.2025.111486","DOIUrl":"10.1016/j.anucene.2025.111486","url":null,"abstract":"<div><div>High-fidelity and high-precision nuclear-thermal coupling calculations for reactors can more accurately simulate the core behavior of a reactor. This work investigates the coupling of neutron physics and thermal–hydraulic behavior using the pin-by-pin subchannel thermal–hydraulic simulation code KMC-FBc based on precise geometric modeling and high-accuracy MOC (Method of Characteristics) neutron transport calculations. To examine the transient characteristics of natural circulation lead–bismuth fast reactors, this work integrates the system code RELAP5 and develops the OpenMOC/KMC-FBc/RELAP5. And the correctness of the coupling program is proved by comparing the calculation results of RELAP5. In this paper, the OpenMOC/KMC-FBc/RELAP5 coupling program is used to simulate the thermal and hydraulic phenomena under the steady state and accident conditions of the natural circulation lead cooled fast reactor (SNCLFR-100) designed by the University of Science and Technology of China. The accident conditions include unprotected transient overpower accident and unprotected loss of heat-sink accident. The results show that the OpenMOC/KMC-FBc/RELAP5 provides high accuracy and can effectively capture the physical and thermal–hydraulic variations in the reactor core under transient conditions.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"219 ","pages":"Article 111486"},"PeriodicalIF":1.9,"publicationDate":"2025-04-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143844188","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":"Comparison of high-fidelity transport code predictions against the NECTAR intra-pin reaction rate measurements","authors":"Tom Mager ,&nbsp;Mathieu Hursin ,&nbsp;Kyle Vaughn ,&nbsp;Sooyoung Choi ,&nbsp;Brendan Kochunas ,&nbsp;Thomas Ligonnet ,&nbsp;Vincent Lamirand ,&nbsp;Oskari Pakari ,&nbsp;Carlo Fiorina ,&nbsp;Andreas Pautz","doi":"10.1016/j.anucene.2025.111430","DOIUrl":"10.1016/j.anucene.2025.111430","url":null,"abstract":"<div><div>The advancement of high-fidelity nuclear reactor simulation methods is essential for reducing conservatism in safety margins and optimizing reactor design and operation. This work evaluates the predictive capabilities of the deterministic solvers GeN-Foam and MPACT for modeling intra-pin reaction rate distributions in heterogeneous reactor cores, using experimental data from the NECTAR campaign conducted at the CROCUS reactor. Radial and azimuthal reaction rate distributions from GeN-Foam and MPACT are compared to NECTAR measurements, with Serpent2 Monte Carlo simulations supporting the validation. GeN-Foam predictions align with measurements within the 0.4% experimental uncertainty, while MPACT results require further improvements to account for the 3D nature of the experiments, particularly in modeling the axial leakage in the vicinity of the dosimeter; and in the cross-section self-shielding treatment of dosimetry materials. To support broader validation efforts, this study provides detailed descriptions of the NECTAR experiments, which are planned to be integrated into the International Reactor Physics Evaluation Project (IRPhEP). Recommendations are made for future validation efforts, including additional measurements with alternative dosimeter materials and enhancements to solver methodologies.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"219 ","pages":"Article 111430"},"PeriodicalIF":1.9,"publicationDate":"2025-04-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143844187","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":"Reactivity measurement using pulsed neutron source methods on VR-1 and VR-2 reactors and their verification by SERPENT code","authors":"Michal Herčík,&nbsp;Jan Rataj,&nbsp;Ondřej Huml,&nbsp;Filip Fejt","doi":"10.1016/j.anucene.2025.111432","DOIUrl":"10.1016/j.anucene.2025.111432","url":null,"abstract":"<div><div>A series of pulse experiments were carried out on the VR-1 reactor and the newly constructed subcritical VR-2 assembly at the Czech Technical University (CTU) in Prague. The objective was to compare the results obtained from the pulse methods with the verified Monte Carlo (Serpent) model for VR-1 reactor and to investigate the potential for the process of experimental verification of the Monte Carlo model for the new VR-2 reactor. The Simmons-King’s, Sjöstrand’s, Gozani’s, and Garelis–Russell’s methods were applied to data collected from detectors located at different positions in a pulse experiment. Additionally, the Monte Carlo method was employed for the determination of the kinetic parameters essential for the comparison and application of the pulse methods. A very simple Serpent simulation of the Sjöstrand’s method was introduced. Although the used pulse source in some of its parameters (frequency, duty factor) indicated a certain limit of applicability, the expected, relatively good agreement between the experiment and the Monte Carlo model was achieved in the investigated range of reactivities for the VR-1 reactor. Therefore, it can be concluded that pulse methods in this range provide a reasonably reliable means of determining reactivity. The discrepancies observed at the VR-2 subcritical assembly indicate the necessity for a rigorous review of its existing Monte Carlo model.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"219 ","pages":"Article 111432"},"PeriodicalIF":1.9,"publicationDate":"2025-04-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143843520","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}