Annals of Nuclear Energy最新文献

{"title":"Cluster-dynamics study on the evolution of dislocation loops in molybdenum: The role of defect diffusion coefficient and equilibrium concentration","authors":"Teng Xie,&nbsp;Xinhua Yan,&nbsp;Menglu Qin,&nbsp;Lu Sun,&nbsp;Zhenfeng Tong","doi":"10.1016/j.anucene.2025.111468","DOIUrl":"10.1016/j.anucene.2025.111468","url":null,"abstract":"<div><div>A cluster dynamics simulation based on rate theory was conducted to investigate the evolution of irradiation-induced dislocation loops in pure molybdenum (<span><math><mi>Mo</mi></math></span>) irradiated with <span><math><mrow><mn>30</mn><mspace></mspace><mi>keV</mi></mrow></math></span> <span><math><mi>He</mi></math></span>⁺ at <span><math><mrow><mn>673</mn><mspace></mspace><mi>K</mi></mrow></math></span>. The changes in number density and size of dislocation loops were analyzed, revealing rapid saturation of loop density with increasing dose, while the average loop radius grows and stabilizes at a maximum size. During the initial irradiation stage, significant nucleation leads to a rapid increase in loop density. The simulation results align well with in-situ transmission electron microscopy observations and clarify the kinetic mechanisms underlying key experimental features. It was found that the diffusion coefficients and thermal equilibrium concentrations of interstitial atoms and vacancies play critical roles in determining the evolution of loop density and size at elevated temperatures. This study provides insight into the microstructural evolution of irradiated molybdenum under helium ion exposure.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"219 ","pages":"Article 111468"},"PeriodicalIF":1.9,"publicationDate":"2025-04-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143863454","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":"Transport-corrected flux-moment homogenization method for generating P0 multigroup cross-section based on continuous energy Monte-Carlo","authors":"Yuyang Shen ,&nbsp;Yiwei Wu ,&nbsp;Qufei Song ,&nbsp;Kuaiyuan Feng ,&nbsp;Hui Guo ,&nbsp;Hanyang Gu","doi":"10.1016/j.anucene.2025.111448","DOIUrl":"10.1016/j.anucene.2025.111448","url":null,"abstract":"<div><div>The scattering anisotropy multigroup cross-section (MGXS) with respect to the outgoing angle, as well as the anisotropy of the total MGXS with respect to the incident angle, are critical for generating high-precision MGXS used in transport core calculations. A TC-MHT method based on continuous energy Monte-Carlo is developed to generate high-accuracy P₀ MGXS. This method addresses scattering anisotropy using transport correction (TC) and handles the anisotropy of total cross-section through the flux-moment homogenization technique (MHT). This method is validated using the Megapower benchmark. In reactivity calculations, the bias is decomposed and the reasons for the bias are analyzed. The bias is reduced to less than 160 pcm by using the TC-MHT method. For calculations of control drum worth, the bias does not exceed 0.2%. In power calculations, the maximum pin-by-pin power distribution deviation is within 4%. This demonstrates that TC-MHT method can generate high-precision, highly adaptable MGXS for deterministic code.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"219 ","pages":"Article 111448"},"PeriodicalIF":1.9,"publicationDate":"2025-04-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143860726","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":"Characteristics analysis of the typical vehicle-mounted megawatt-scale heat pipe cooled reactor power system","authors":"Henglong Lin ,&nbsp;Zeguang Li ,&nbsp;Kan Wang ,&nbsp;Weilin Zhuge ,&nbsp;Ganglin Yu ,&nbsp;Hongsheng Jiang ,&nbsp;Jingkang Li ,&nbsp;Jun Yang ,&nbsp;Zilin Su ,&nbsp;Huifu Wang","doi":"10.1016/j.anucene.2025.111467","DOIUrl":"10.1016/j.anucene.2025.111467","url":null,"abstract":"<div><div>Heat pipe cooled reactor (HPR) has the characteristics of small size and high safety, and can be used in special application scenarios such as vehicle-mounted mobility. In this paper, the Mobile Nuclear Power System based on HPR named MNPS-1000, which is designed to generate 1MWe, is selected for research and analysis. According to the design, a whole system model is established, which includes the reactor model, high-temperature heat pipe model, heat pipe heat exchanger model and energy conversion system model, etc. Using MATLAB/SIMULINK, the analysis platform suitable for this nuclear power system is built. Based on this platform, the steady state and transient performance of the system are characterized. In the steady state analysis, the simulated values of key nodes of the model are in general agreement with the calculated values of parameter matching, and the maximum relative error does not exceed 6 %. The effectiveness of characteristic analysis based on the built platform is verified. In the transient analysis, typical reactor system accidents and energy conversion system accident are simulated and analyzed respectively. The maximum temperature of the fuel assemblies in these accidents does not exceed 1550 K, which is lower than the selected material temperature safety limit. The inherent safety feature of the system is discussed.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"219 ","pages":"Article 111467"},"PeriodicalIF":1.9,"publicationDate":"2025-04-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143860309","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":"Neutron resonance integral analysis of 238U within resolved resonance region","authors":"Yufei Pan,&nbsp;Jimin Wang,&nbsp;Ruirui Xu,&nbsp;Xi Tao,&nbsp;Yue Zhang,&nbsp;Jie Ren,&nbsp;Yuan Tian,&nbsp;Zhigang Ge,&nbsp;Nengchuan Shu,&nbsp;Kang Xing,&nbsp;Xiaofei Wu","doi":"10.1016/j.anucene.2025.111450","DOIUrl":"10.1016/j.anucene.2025.111450","url":null,"abstract":"<div><div>Neutron resonance data are significant importance for applications in nuclear engineering. Aimed to assess the discrepancies efficiently between the experimental measurements and evaluations, including ENDF/B-VIII.1, JEFF-3.3, JENDL-5, CENDL-3.2 and BROND-3.1. A resonance peak identification technique based on derivative method and Gaussian Filter is established to conduct systematic analysis on two resonance properties (integral values and central energy) of the experimental and evaluated data within the resolved resonance region. The deviations of the two properties of ²³⁸U are analyzed in this work. More than 1500 resonance peaks of neutron total and capture reactions are identified below neutron energy 20 keV. As a result, the evaluated central energies of (n, tot) resonances exhibit excellent agreement with experimental data, while those of (n, <span><math><mi>γ</mi></math></span>) resonances show increasing deviation with rising energy levels; significant deviations of resonance integrals exist between experimental and evaluated data of (n, <span><math><mi>γ</mi></math></span>), and (n, tot) demonstrate a better consistency. The current work efficiently identifies the energy regions that require improvement in the future.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"219 ","pages":"Article 111450"},"PeriodicalIF":1.9,"publicationDate":"2025-04-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143860737","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 of internal flow induced vibration of tray rod assembly during on-power handling in a typical Indian research reactor","authors":"A. Ravi Kiran ,&nbsp;M.K. Agrawal ,&nbsp;Renuka Patel ,&nbsp;S.K. Sinha","doi":"10.1016/j.anucene.2025.111511","DOIUrl":"10.1016/j.anucene.2025.111511","url":null,"abstract":"<div><div>A tray rod assembly in a research reactor is a facility that irradiates samples inside reactor core. For on-power lifting of the tray rod assembly, it is necessary to ensure its integrity under axial flow induced vibrations, resulting from coolant flow. In the present work, methodology for qualifying the axial flow induced vibration response of a tray rod assembly is demonstrated. Experiments are conducted to study the axial flow induced vibration response of an ‘on-power tray rod’ of a typical Indian Research Reactor (IRR), using an in-house flow test facility. Three conditions are proposed for the qualification of ‘on-power’ tray rod assembly under axial flow induced vibration load. First one is the permissible vibration velocity, second one is the permissible displacement and the third one is the permissible flow velocity. Free vibration analysis of the assembly is carried out to obtain the dynamic characteristics and to compare with experimental results.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"219 ","pages":"Article 111511"},"PeriodicalIF":1.9,"publicationDate":"2025-04-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143860727","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":"Optimization of heterogeneous parallel algorithm for Monte Carlo neutron transport simulation aiming at thread divergence Issues","authors":"Xian Zhang ,&nbsp;Shu Li ,&nbsp;Xin Wang ,&nbsp;Danhua ShangGuan ,&nbsp;Shichang Liu","doi":"10.1016/j.anucene.2025.111508","DOIUrl":"10.1016/j.anucene.2025.111508","url":null,"abstract":"<div><div>The Monte Carlo simulation of large-scale neutron transport problems has always faced the problem of slow computation. In order to fully exploit the acceleration advantage of heterogeneous parallelism on the Monte Carlo neutron transport simulation, this paper carries out research around the history-based neutron tracking algorithm, deeply explores the adaptation of the Monte Carlo algorithm and heterogeneous parallelism. Aiming at the thread divergence problem, optimization strategies for particle tracking algorithm are proposed to ensure load balancing among parallel threads. In addition, to mitigate the impact of global memory access latency, the memory layout of the particle state data is reasonably arranged by comprehensively considering the random memory access of Monte Carlo algorithm and the hardware characteristics of GPU. The reliability and efficiency of heterogeneous parallel algorithm are validated in calculations of benchmarks, the computing performance on an NVIDIA A800 GPU is equivalent to the performance of 62–87 CPU cores.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"219 ","pages":"Article 111508"},"PeriodicalIF":1.9,"publicationDate":"2025-04-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143863598","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":"Scale model for investigating two-phase flow instability in a PWR passive residual heat removal system","authors":"Jong Chull Jo ,&nbsp;Frederick J. Moody","doi":"10.1016/j.anucene.2025.111493","DOIUrl":"10.1016/j.anucene.2025.111493","url":null,"abstract":"<div><div>To design a scale model of a PWR passive residual heat removal system that preserves and controls the two-phase flow instability, defined as unstable oscillations of the condensate water level at the saturated steam/water interface in the vertical condensate pipe column, non-dimensional (Pi) groups were derived using the top-down (macroscopic) modeling approach known as the <em>Buckingham Pi Theorem</em>. Subsequently, the scaling ratios and scaling laws for the two-phase flow instability-related parameters of the system design—expressed in terms of the ratios of pipe length and diameter between the scale model and the full-size system—were derived based on the relationships between the Pi groups. The scaling ratios obtained for the parameters such as length, diameter, area, volume, time, oscillation frequency, mass flow rate, heat input, and pressure drop were compared with those derived from bottom-up (microscopic/fine structure) modeling approaches, including the three-level and hierarchical two-tiered scaling methods. As a result, the findings of this study align with those obtained using either the three-level or hierarchical two-tiered scaling method, except for the heat input (power) scaling ratio and the power-volume scaling ratio. From a dimensional similarity perspective, the present method appears to yield physically consistent values for the heat input and power-volume scaling ratios, which are defined as the product of the length scaling ratio and the corresponding scaling ratio derived from either the three-level or hierarchical two-tiered scaling method. Consequently, this study makes a significant contribution to engineering literature by demonstrating that the <em>Buckingham Pi Theorem</em> can establish scaling laws for designing a scale model to predict and interpret full-scale system behavior without requiring a comprehensive understanding of transient two-phase flow phenomena, unlike bottom-up approaches that necessitate detailed knowledge of the underlying phenomena.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"219 ","pages":"Article 111493"},"PeriodicalIF":1.9,"publicationDate":"2025-04-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143855053","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 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}