Annals of Nuclear Energy最新文献

{"title":"Numerical study on three-dimensional coupling characteristics for electromagnetism and flow in flat linear induction pump","authors":"Heyao Li ,&nbsp;Zhuqian Zhang ,&nbsp;Lixin Yang ,&nbsp;Keming Bi ,&nbsp;Xin Yan ,&nbsp;Haojie Zhang ,&nbsp;Xiao Chen","doi":"10.1016/j.anucene.2025.111876","DOIUrl":"10.1016/j.anucene.2025.111876","url":null,"abstract":"<div><div>Electromagnetic pump (EMP) is a device that transports liquid metal by Lorentz force generated from the vector product of the magnetic induction intensity and eddy current density in liquid metal. In this paper, a three-dimensional magnetohydrodynamic (MHD) numerical model is constructed for a flat linear induction pump (FLIP). A coupling model between the electromagnetic model and the flow model is proposed by inserting the instantaneous Lorentz force fields as the momentum source term into Navier-Stokes equations (N-S equations) in the form of a segmented function. The characteristics of the temporal and spatial distribution of Lorentz force field on the NaK78 fluid under the joint action of magnetic field and electric field are numerically analyzed. Based on the influence of the temporal and spatial distribution of Lorentz force field, the flow irregularity of liquid metal and the fluctuation characteristics of head in FLIP under the varying flow rate are investigated. In order to accommodate the rapid verification for the design of FLIP, an equivalent coupling model based on changing the form of the Lorentz force field distribution by exploring the influence of the homogeneity of the distribution in both time and space on the head characteristics of FLIP is developed. The accuracy of the MHD numerical model and the equivalent coupling model are verified by comparing with the experimental results of the head that obtained from a principle prototype loop test system of FLIP actually constructed.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"226 ","pages":"Article 111876"},"PeriodicalIF":2.3,"publicationDate":"2025-09-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145099854","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":"Reliability analysis of impeller and shaft system in shaft-sealed reactor coolant pump","authors":"Long Cai ,&nbsp;Zhe Jiao ,&nbsp;Dongbo Qi ,&nbsp;Xuezhi Ding ,&nbsp;Yuan Xu ,&nbsp;Yun Long","doi":"10.1016/j.anucene.2025.111884","DOIUrl":"10.1016/j.anucene.2025.111884","url":null,"abstract":"<div><div>This study addresses the reliability issues of the axial seal main pump (RCP) impeller and shaft system under high-temperature and high-pressure water conditions. It employs fluid–structure interaction (FSI) theory and modal analysis to conduct a systematic study of dynamic characteristics and fatigue life. Using FSI simulation model, this study compares the natural frequency variations under dry and wet modal conditions to reveal the influence of added water mass on structural dynamics. Transient dynamic stress calculations are completed on the ANSYS Workbench platform. The fatigue life of the impeller is predicted using Miner’s linear cumulative damage theory and nCode DesignLife software. The results indicate that most regions of the impeller have a lifespan exceeding 7.223 × 10¹⁰ cycles (approximately 92.5 years). Localized fatigue risks are identified only at the blade roots. Campbell diagram analysis shows that the rated speed of the nuclear main pump avoids the main resonant critical speeds, but attention should be paid to the influence of low-frequency excitation on stability. The findings of this study provide a theoretical basis for the structural optimization and reliability enhancement of nuclear main pumps.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"226 ","pages":"Article 111884"},"PeriodicalIF":2.3,"publicationDate":"2025-09-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145099853","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 and numerical research of thermo-hydraulic characteristics of liquid lead–bismuth eutectic and S-CO2 in a MCHE","authors":"Hong Liu ,&nbsp;Zhigang Zhang ,&nbsp;Shuo Yang ,&nbsp;Guangliang Chen ,&nbsp;Zhibin Rao ,&nbsp;Runze Wang","doi":"10.1016/j.anucene.2025.111879","DOIUrl":"10.1016/j.anucene.2025.111879","url":null,"abstract":"<div><div>This study utilized experimental methods to investigate the variation of the overall heat transfer coefficient in a microchannel heat exchanger (MCHE) with the Reynolds numbers of liquid lead–bismuth eutectic (LBE) and supercritical carbon dioxide (S-CO₂). Additionally, we obtained numerical values for the resistance of liquid LBE and S-CO₂ with respect to their respective Reynolds numbers. Furthermore, we employed numerical simulation methods to analyze the thermal hydraulics characteristics of LBE and S-CO₂ within the MCHE. We found that the numerical simulation results were in good agreement with the experimental data. Using the numerical results, we further analyzed the variations of the mainstream temperature, wall temperature, and heat flux density for both fluids. We also investigated the influence of fluid velocities and heat exchanger spacing on heat transfer. Finally, we proposed correlation equations for the calculation of heat transfer and resistance for LBE and S-CO₂.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"226 ","pages":"Article 111879"},"PeriodicalIF":2.3,"publicationDate":"2025-09-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145099856","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":"A novel kinetic model for dissolution and precipitation of oxide on stainless-steel surface in stagnant liquid sodium","authors":"Munemichi Kawaguchi ,&nbsp;Asuka Ikeda ,&nbsp;Jun-ichi Saito","doi":"10.1016/j.anucene.2025.111880","DOIUrl":"10.1016/j.anucene.2025.111880","url":null,"abstract":"<div><div>This study presents sodium experiments and a novel kinetic model to elucidate the dissolution and precipitation behavior of oxides on the stainless-steel (SS) surface in stagnant liquid sodium. Experiment results demonstrated that the oxygen from Na₄FeO₃ on the SS surface dissolved into the liquid sodium at a rate of <span><math><mrow><msub><mi>v</mi><mrow><mi>d</mi><mi>i</mi><mi>s</mi></mrow></msub><mo>≈</mo><mn>9.3</mn><mo>×</mo><msup><mrow><mn>10</mn></mrow><mrow><mo>-</mo><mn>5</mn></mrow></msup></mrow></math></span> wt.ratio/h within 20 h. Subsequently, oxide precipitation occurred on the SS surface at a rate of <span><math><mrow><msub><mi>v</mi><mrow><mi>p</mi><mi>r</mi><mi>e</mi></mrow></msub><mo>≈</mo><mn>1.4</mn><mo>×</mo><msup><mrow><mn>10</mn></mrow><mrow><mo>-</mo><mn>5</mn></mrow></msup></mrow></math></span> wt.ratio/h. To further investigate the precipitation dynamics, a phase-field (PF) simulation code was developed, enabling sensitivity analyses of six parameters: temperature (<span><math><mi>T</mi></math></span>), initial oxygen concentration (<span><math><msub><mi>c</mi><mn>0</mn></msub></math></span>), interfacial energy (<span><math><mi>γ</mi></math></span>), oxygen diffusion coefficient (<span><math><msub><mi>D</mi><mi>L</mi></msub></math></span>), quasi-partition coefficient (<span><math><mi>k</mi></math></span>), and time step (<span><math><mrow><mi>Δ</mi><mi>t</mi></mrow></math></span>). The PF simulations revealed that the precipitation rate increases linearly with <span><math><msub><mi>c</mi><mn>0</mn></msub></math></span> and <span><math><msub><mi>D</mi><mi>L</mi></msub></math></span>. In contrast, it decreases exponentially with increasing <span><math><mi>T</mi></math></span> and <span><math><mi>γ</mi></math></span>. The parameters <span><math><mi>k</mi></math></span> and <span><math><mrow><mi>Δ</mi><mi>t</mi></mrow></math></span> exhibited negligible influence on the calculation results. Notably, the oxide precipitation rates obtained by the PF calculation show consistency with the laboratory-scale experimental findings reported by LatgÉ et al, supporting the model’s validity.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"226 ","pages":"Article 111880"},"PeriodicalIF":2.3,"publicationDate":"2025-09-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145099855","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 internal heat generation on turbulent heat transfer in a pipe flow with wall cooling","authors":"Dong-Hyuk Park,&nbsp;Bum-Jin Chung","doi":"10.1016/j.anucene.2025.111886","DOIUrl":"10.1016/j.anucene.2025.111886","url":null,"abstract":"<div><div>We investigated the influence of internal heat generation on turbulent heat transfer in a pipe flow, aiming at application to the heat exchangers of Molten Salt Reactors (MSRs), where nuclear fuel is dissolved in the coolant. The non-homogeneous energy equation arising from the internal heat source was solved using the superposition principle and calculations were conducted from the entrance region to the fully developed region. Local and mean Nusselt numbers (<em>Nu</em>) were evaluated, varying Reynolds numbers (<em>Re</em> = 5 × 10³–10⁶), Prandtl numbers (<em>Pr</em> = 1–10), and internal heat generation parameters (Ω = 1–10³). Results show that the internal heat generation enhances the heat transfer rate. This enhancement increases with higher Ω and lower <em>Re</em> and <em>Pr</em>. However, due to the thin viscous sublayer in turbulent flow, the maximum local <em>Nu</em> enhancement remained below 12 %. Flow regime where internal heat generation produces a significant impact were identified, and a correction factor was developed. This study provides fundamental insights into the influence of internal heat generation and offers a quantitative basis for the design of MSR heat exchangers.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"226 ","pages":"Article 111886"},"PeriodicalIF":2.3,"publicationDate":"2025-09-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145099849","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":"Numerical study of lead oxide precipitation in liquid lead–bismuth eutectic","authors":"Jinye Luo ,&nbsp;Maolong Liu ,&nbsp;Xianglong Guo","doi":"10.1016/j.anucene.2025.111885","DOIUrl":"10.1016/j.anucene.2025.111885","url":null,"abstract":"<div><div>Liquid lead–bismuth eutectic (LBE) is considered a promising primary coolant for Generation IV nuclear fast reactors. However, the ingress of oxygen into LBE cooled reactors can cause lead oxide (PbO) precipitation. In order to evaluate the safety in the presence of PbO, quantification study of PbO precipitation in LBE cooled reactor especially in heat exchangers is necessary while insufficient study is currently unveiled. In this study, the nucleation, growth and dissolution of PbO are concerned. The population balance model is applied to characterize the distribution of particle sizes. With the use of CFD code ANSYS FLUENT, the simulation model was validated using the existing data. Then, the model was applied to the calculations of Printed Circuit Heat Exchanger (PCHE). The amount and location of PbO precipitates in PCHE under various conditions were calculated. The numerical simulation results are useful to improve the safety and performance of LBE cooled reactors.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"226 ","pages":"Article 111885"},"PeriodicalIF":2.3,"publicationDate":"2025-09-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145099852","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":"Uncertainty quantification of a physics-informed model based on sparse identification of a Thermal Energy Distribution System","authors":"Paul Seurin,&nbsp;Linyu Lin","doi":"10.1016/j.anucene.2025.111865","DOIUrl":"10.1016/j.anucene.2025.111865","url":null,"abstract":"<div><div>Integrated energy systems (IES)s are crucial for enhancing the economy and efficiency of power generation sources (e.g., nuclear energy) necessary to unleash American energy dominance. These systems can be integrated with thermal energy storage (TES) and intermittent renewable energies to optimize overall energy use, peak-load regulation, and demand-side responses. However, the stabilization of energy generation, transport, and utilization introduces operational complexities that exceed the challenges of managing each sub-component individually. Currently, though IESs rely on human operators for efficiency and stability, reducing human error risk and enhancing performance through automation is highly desirable. Recent advances at Idaho National Laboratory have demonstrated successful control of the Thermal Energy Distributed System (TEDS). However, the automatic control system depends on a deterministic Sparse Identification of Nonlinear Dynamics with Control (SINDyC) model, which are trained based on simulation data from physics-based simulations. Because of uncertainties in physics-based simulation, SINDyC model results in large discrepancies against experimental data and cannot be reliably used in automatic control. In this paper, we present an innovative approach to address these discrepancies by quantifying uncertainties and developing a more robust model. We first generated trajectories by using first-principles physics codes to encapsulate the experiment. Next, we trained thousands of models by randomly sampling these trajectories. We then collapsed all those models into one probabilistic SINDyC by fitting a multivariate Gaussian distribution onto the resulting coefficient’s distribution. Despite its simplicity, our approach successfully produced 95% confidence intervals that captured the experimental trajectories. It even did so with a higher probability and better U-pooling score across six of the seven relevant quantities of interest (QoIs), as compared to other classical approaches. Ongoing research is focusing on generating new experimental trajectories to validate this approach, and on employing Bayesian calibration to refine parametric uncertainties and guide future model development efforts.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"226 ","pages":"Article 111865"},"PeriodicalIF":2.3,"publicationDate":"2025-09-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145099851","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":"Integrated items safety classification method considering risk evaluation for nuclear reactors","authors":"Ming Sun ,&nbsp;Xiaoer He ,&nbsp;Longlong Tao ,&nbsp;Shaoxuan Wang ,&nbsp;Daochuan Ge ,&nbsp;Chenglong Cao ,&nbsp;Zhongchang Zhang","doi":"10.1016/j.anucene.2025.111877","DOIUrl":"10.1016/j.anucene.2025.111877","url":null,"abstract":"<div><div>The safety classes of items, which are the basis of selecting appropriate design and manufacture codes, seismic grades, and quality assurance classes, are directly related to the safety and economy of nuclear reactors. The reasonable safety classification of items is very important for improving the safety and economy of nuclear reactors. The current safety classification of items is sometimes too conservative, without combining different safety classification approaches, such as function based, consequence based, and probability based. This contribution firstly introduced and compared the 50-SG-D1, SSG-30, and risk-informed safety classification methods briefly. Then, based on these safety classification methods, an integrated items safety classification method considering risk evaluation for nuclear reactors was proposed, and a case study was performed on the Vacuum Vessel Pressure Suppression System (VVPSS) of International Thermonuclear Experimental Reactor (ITER) to demonstrate the effectiveness of the proposed method. The results show that the proposed method, which fully integrates the characteristics of 50-SG-D1, SSG-30, and risk-informed safety classification methods, is more comprehensive and advanced.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"226 ","pages":"Article 111877"},"PeriodicalIF":2.3,"publicationDate":"2025-09-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145098673","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 flow characteristics on thermal energy storage during Paraffine wax melting process","authors":"Adnan Muhammad ,&nbsp;Hui Wang ,&nbsp;Wang Yanlu ,&nbsp;Khan Jamil ,&nbsp;Cao Xiaxin","doi":"10.1016/j.anucene.2025.111872","DOIUrl":"10.1016/j.anucene.2025.111872","url":null,"abstract":"<div><div>This study investigates the thermal performance of paraffin-based phase change material (PCM) in a Passive Containment Cooling System (PCCS), specifically for nuclear safety applications. Using validated numerical simulations, the effects of key parameters, including inlet temperature (100–140 °C), Thermal stratification of PCM domain, velocity of high temperature fluid (0.3–0.9 m/s), flow direction (bottom-to-top vs. top-to-bottom) of HTF, and heating medium (water vs. steam) on PCM melting dynamics were systematically analyzed. The results reveal that higher inlet temperatures and flow velocities significantly reduce melting Time and improve heat transfer coefficients, thereby enhancing energy absorption efficiency. Among the flow configurations, top-to-bottom flow outperforms bottom-to-top flow due to improved convective interaction. At the same Time, steam heating proves more effective than hot water by utilizing latent heat. These findings emphasize the critical role of optimizing thermal and flow parameters in improving the performance of PCM-based thermal energy storage systems for emergency cooling in nuclear reactors. Future research will focus on advanced heat transfer designs and enhancing the long-term reliability of PCM-based systems for nuclear safety applications.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"226 ","pages":"Article 111872"},"PeriodicalIF":2.3,"publicationDate":"2025-09-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145098672","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":"Variation of primordial radionuclide activities in the drinking water collected from Darbandikhan Lake water resources during four seasons","authors":"Adeeb Omer Jafir,&nbsp;Ali Hassan Ahmed","doi":"10.1016/j.anucene.2025.111883","DOIUrl":"10.1016/j.anucene.2025.111883","url":null,"abstract":"<div><div>The variation in the levels of natural radioactive materials in drinking water throughout the four seasons has been studied in Darbandikhan Lake using a gamma spectrometer system with an HPGe detector for the measurements. The arithmetic activity concentrations for ²²⁶Ra, ²³²Th and ⁴⁰K in the water samples were established to be 1.54 ± 0.21, 2.39 ± 0.0.33 and 77.94 ± 1.70 Bq/L; 1.22 ± 0.18, 2.38 ± 0.31 and 63.79 ± 1.48 Bq/L; 1.18 ± 0.18, 1.89 ± 0.30 and 51.30 ± 1.33 Bq/L; 1.34 ± 0.20, 2.30 ± 0.32 and 68.14 ± 1.56 Bq/L, for spring, summer, autumn, and winter, respectively. It was observed that ²³²Th was about twice as high as the WHO’s recommended levels for drinking water. In wet seasons of spring and winter, Drinking water show higher average activity concentration than those in summer and autumn. The results obtained from ANOVA statistical tests showed that only ⁴⁰K had a statistically significant difference in concentration throughout the four seasons (p = 0.029). Both the external and internal hazard indices (0.025 and 0.03) are significantly below the recommended value of one. The committed effective dose is measured at 0.70 mSv/y, while the annual effective doses for outdoor and indoor exposure are 0.006 mSv/y and 0.043 mSv/y, respectively. The calculated excess lifetime cancer risk due to water ingestion is 2.46 × 10⁻³, whereas the cancer risk for indoor and outdoor exposure is 0.02 × 10⁻³ and 0.15 × 10⁻³, respectively. With the exception of the annual committed effective dose and the cancer risk associated with water consumption, all hazard indices are significantly below the recommended global standards.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"226 ","pages":"Article 111883"},"PeriodicalIF":2.3,"publicationDate":"2025-09-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145098674","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}