Annals of Nuclear Energy最新文献

{"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":"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}