Progress in Nuclear Energy最新文献

{"title":"Numerical analysis of heat transfer characteristics in the nuclear waste storage vaults based on orthogonal experiment","authors":"Yu Zhao ,&nbsp;Heng Zhou ,&nbsp;Seyed Majid Bigonah Ghalehsri ,&nbsp;Juan Xu ,&nbsp;Jianfu Zhang ,&nbsp;Tianyi Zhao ,&nbsp;Jiantao Shao","doi":"10.1016/j.pnucene.2025.105853","DOIUrl":"10.1016/j.pnucene.2025.105853","url":null,"abstract":"<div><div>The decay process of nuclear waste in storage tanks generates a large amount of heat, and the uninterrupted operation of the ventilation system in the repository wastes a lot of energy. This study focuses on analyzing the change rule of temperature distribution with 1/3 amount of the total capacities inside a typical nuclear waste storage repository, taking into account different tank arrangements, outlet locations and air intake angles, etc. The method of orthogonal design of experiments was used to determine 16 computational fluid dynamics (CFD) simulation cases, in which heat conduction, convection, and radiation are included in the simulation, and to establish the maximum temperature of the wall surface, the average temperature of the surface of the heat source, and the average temperature in the space according to the dimensionless temperatures were evaluated, The results showed that none of the wall surface temperatures of the 16 cases exceeded the temperature limit of 65 °C, the wall surface temperatures of the storage tanks in a centralized row were 16 % higher than those in a decentralized row, and the heat transfer was best in the case of staggered rows, with air outlets located on both sides, and with vertical inflow. This study could be the guidance for the design of nuclear waste storage repository, which provides ideas for the optimization scheme during operation, and improves the efficiency of energy use under the premise of ensuring safety.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"187 ","pages":"Article 105853"},"PeriodicalIF":3.3,"publicationDate":"2025-05-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144179471","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":"Implementation of dry storage models in Falcon: cladding creep, annealing of cladding irradiation damage, and fuel helium-swelling","authors":"C. Cozzo ,&nbsp;P. Konarski ,&nbsp;G. Khvostov ,&nbsp;B. Mervin ,&nbsp;H. Ferroukhi","doi":"10.1016/j.pnucene.2025.105859","DOIUrl":"10.1016/j.pnucene.2025.105859","url":null,"abstract":"<div><div>Dry storage is an important phase of the nuclear fuel cycle as it allows a reduction in the decay heat so that the used fuel can be safely transported to and stored in a final repository. However, the dry storage period lasts several decades and the integrity of the fuel rods must be guaranteed during this time. The fuel performance code Falcon allows simulating nuclear rods in reactor operating conditions, but the models are not taking into account the dry storage conditions which have a greater time scale and different thermal and mechanical environments. The introduction of specific models in Falcon is necessary for the reliable simulation of fuel rods in dry storage. Therefore, new cladding creep, cladding annealing and fuel helium-swelling models have been implemented in the code and the validation is performed for each model. Base irradiation simulations have been performed for a Swiss PWR representative rod and a BWR rod. Dry storage conditions from a benchmark have been applied. Gap closure is identified as the main contributor to stress increase in the cladding. In the BWR case, no gap closure is calculated and little difference is observed in the cladding mechanical response for different models. In the PWR case, gap closure is calculated and it is observed that the annealing model and cladding temperature history have a limited effect on the cladding hoop stress. On the other hand, the fuel helium-swelling model has a pronounced impact on the calculated stresses. In realistic conditions and with the use of the dry storage models, the calculated hoop stress is below the critical values of 90 and 70 MPa for PWR and BWR, respectively. Also, the calculated Hoop strain is below 1 % in realistic temperature conditions.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"187 ","pages":"Article 105859"},"PeriodicalIF":3.3,"publicationDate":"2025-05-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144169912","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":"Optimizing exergy efficiency in the Bushehr WWER-1000 nuclear power plant using Multi-Verse Optimization algorithm","authors":"M. Safari,&nbsp;M. Aghaie,&nbsp;H. Ebrahimgol,&nbsp;S. Ghiasabadi","doi":"10.1016/j.pnucene.2025.105849","DOIUrl":"10.1016/j.pnucene.2025.105849","url":null,"abstract":"<div><div>Exergy analysis and optimization methods are essential for evaluating thermal energy processes in power plants. This study aims to optimize the thermal efficiency of a nuclear power plant (NPP) using the exergy method. In this work, the energy and exergy analysis of the Bushehr Nuclear Power Plant (BNPP) are assessed, followed by an exploration of various optimization algorithms in conjunction with the first and second laws of thermodynamics. By employing advanced computational techniques, overall plant performance is improved. Using algorithms such as Multi-Verse Optimization (MVO), Arithmetic Optimization Algorithm (AOA), Bird Swarm Algorithm (BSA), Salp Swarm Algorithm (SSA), Marine Predators Algorithm (MPA), Flow Direction Algorithm (FDA), and Whale Optimization Algorithm (WOA), the thermal efficiency of the BNPP is improved from 33.6 % to 36.44 %, 36.42 %, 35.486 %, 35.48 %, 35.42 %, 35.464 %, and 35.385 %, respectively. The average results over ten runs indicate an exergy ranking of MVO &lt; AOA &lt; BSA &lt; SSA &lt; MPA &lt; FDA &lt; WOA. Results show that main equipment, such as the reactor, steam generator, and reactor coolant pumps, exhibited constant irreversibility. In contrast, other components, including the low-pressure turbine (LPT), re-heater, deaerator, condensate pump, and High-Pressure Heater 1 (HPH1), demonstrated increasing in irreversibility, and totally the irreversibility was reduced entire the plant.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"187 ","pages":"Article 105849"},"PeriodicalIF":3.3,"publicationDate":"2025-05-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144169911","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 refined technique for extracting strontium from water samples and quantifying strontium radioisotopes using a gas flow proportional counter","authors":"Phan Long Ho ,&nbsp;Vu Tuan Minh ,&nbsp;Le Dinh Hung ,&nbsp;Le Hoang Minh ,&nbsp;Tran Thien Thanh ,&nbsp;Dang Van Chinh ,&nbsp;Chau Van Tao","doi":"10.1016/j.pnucene.2025.105856","DOIUrl":"10.1016/j.pnucene.2025.105856","url":null,"abstract":"<div><div>This study focuses on developing an improved analytical method for determining strontium radioactivity (Sr-89 and Sr-90) in water samples. While the original procedure used high concentrations of acid, requiring more sample processing time and reducing safety and analytical performance, the proposed procedure addresses all the drawbacks of the old method. The new method optimizations include using 65 % HNO₃ instead of 90 % HNO₃ for safety, reducing separation steps from forty-three to twenty-six to save time and costs, and shortening the century equilibration between Sr-90 and Y-90 to under two weeks. Method validation on sixty-one samples shows the detection limit is 0.035 Bq.L⁻¹, meeting EPA requirements of 0.074 Bq.L⁻¹. Strong linearity (r = 0.999) between spiked and measured activity concentrations was observed in the working range of 0.046–23.371 Bq.L⁻¹ encompasses WHO's 10 Bq.L⁻¹ limit for Sr-90 in drinking water. Interlaboratory testing of three IAEA samples over 2021–2023 further verifies accuracy. The results indicated that the new method has improved efficiency and safety. That reduced costs and analysis time, demonstrating the method's suitability for various radioanalytical applications, particularly emergency nuclear scenarios demanding rapid Sr-90 isotope quantification.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"187 ","pages":"Article 105856"},"PeriodicalIF":3.3,"publicationDate":"2025-05-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144169783","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 multi-scale model of equivalent irradiation creep rate for U-10Mo/Zr dispersion fuels during irradiation","authors":"Yong Li ,&nbsp;Jing Zhang ,&nbsp;Xiaobin Jian ,&nbsp;Guochen Ding ,&nbsp;Yunxuan Cai ,&nbsp;Feng Yan ,&nbsp;Shurong Ding ,&nbsp;Yuanming Li","doi":"10.1016/j.pnucene.2025.105857","DOIUrl":"10.1016/j.pnucene.2025.105857","url":null,"abstract":"<div><div>Irradiation creep is a critical factor influencing the performance and safety of nuclear fuels. The macroscopic irradiation creep response of dispersion fuels is primarily dictated by the mesoscopic irradiation-induced behaviors of fuel particles and the inert matrix, including irradiation creep, swelling, and microstructure evolution. In this study, a finite element (FE) simulation of irradiation creep tests was conducted for U-10Mo/Zr dispersion fuels using a representative volume element (RVE) model with randomly dispersed U-10Mo particles in a Zr matrix. A multi-scale model for the equivalent irradiation creep rate was established based on the analysis of these behaviors. The key findings are: (1) The average deviatoric stress of U-10Mo particles decreased nonlinearly and stabilized during irradiation creep tests due to porosity increase induced by fission gas bubble growth; (2) A parameter combining average deviatoric stress and porosity remained almost constant, simplifying the modeling of the equivalent irradiation creep strain rate; (3) The results from the multi-scale model matched well with the FE simulation results, thereby verifying the accuracy and applicability of the model. This study establishes a novel multi-scale modeling approach, providing a fundamental model for predicting irradiation-induced thermo-mechanical behaviors in fuel elements and assemblies. The model accurately captures the dynamic interplay between microstructural evolution and macroscopic mechanical response, offering significant insights for the design and optimization of advanced nuclear fuels.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"187 ","pages":"Article 105857"},"PeriodicalIF":3.3,"publicationDate":"2025-05-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144169782","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":"Elastoplastic fracture of Cr-coated accident tolerant fuel cladding under multi-field coupling","authors":"Xin Wang ,&nbsp;Ze Xu ,&nbsp;Yulan Liu ,&nbsp;Biao Wang","doi":"10.1016/j.pnucene.2025.105860","DOIUrl":"10.1016/j.pnucene.2025.105860","url":null,"abstract":"<div><div>In this paper, the Cr-coated accident tolerant fuel (ATF) cladding model with initial crack is developed to simulate mechanical behaviors when considering the multi-field coupling. Initial cracks with different lengths and inclination angles in coating are detailed analyzed by utilizing the finite element method. The stress near crack tip and the extents of plastic zones are contrasted. Moreover, two failure modes of coating, which contains 1) failure caused by crack propagation and 2) failure due to the penetration of the plastic zone, are determined. Corresponding failure conditions are also clarified. Results show that the increase of crack length and inclination angle may raise stress levels near the crack tip and lead to accumulation of plastic strain at the junction of substrate and coating, further leading the variation on plastic zone. Cracks with a smaller length and a larger inclination angle possess higher stability. Cracks shorter than approximately 27.1 μm are stable, maintaining the coating's functionality. Results and Conclusions in this paper can provide reference and guidance for future experiments, especially can help improve Cr-coated ATF cladding design and guide future research on factors like crack density and coating types for better protection.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"187 ","pages":"Article 105860"},"PeriodicalIF":3.3,"publicationDate":"2025-05-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144169784","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":"Enhancing nuclear security frameworks: A review of methods and tools","authors":"Md Shafiqul Islam,&nbsp;Zareen Tahsin Anjum,&nbsp;Tasfia Rahman Riva","doi":"10.1016/j.pnucene.2025.105852","DOIUrl":"10.1016/j.pnucene.2025.105852","url":null,"abstract":"<div><div>In the realm of nuclear security research and education, access to appropriate methodologies, tools, and resources remains a critical focal point. While nuclear safety research methods, tools, and education have gained widespread recognition and institutional support across academia, research organizations, and technical support entities, nuclear security continues to face limited visibility, accessibility, and prioritization. This disparity creates challenges for educators, researchers, and professionals seeking effective methods, tools, and cooperation to address nuclear security threats. Given that nuclear security is equally vital as nuclear safety in ensuring the safe and responsible use of nuclear energy, neglecting it poses serious implications, compromising nuclear security inherently undermines nuclear safety. Furthermore, the development and deployment of advanced reactors may introduce new security vulnerabilities that require novel approaches to threat analysis and response. This study reviews a range of nuclear security methodologies and tools documented in the literature, aiming to familiarize stakeholders with their strengths, limitations, and practical applications. By doing so, it supports the advancement of nuclear security education and research. The paper highlights the need for stakeholders including educators, researchers, and practitioners to identify and apply context-appropriate methods and tools tailored to specific facility types and scenarios. In doing so, the study also uncovers several obstacles, including confidentiality, the limited availability of open-source tools, gaps in research and educational programs, and a shortage of career opportunities. The study concludes by offering recommendation for policymakers to address these gaps and promote a more robust, inclusive, and sustainable nuclear security education and research landscapes.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"187 ","pages":"Article 105852"},"PeriodicalIF":3.3,"publicationDate":"2025-05-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144169781","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 and analysis of swirling flow regime transition and separation performance for a new cycloidal vane separator at high gas-liquid velocity ratios","authors":"Luyun Mao,&nbsp;Nan Gui,&nbsp;Yanfei Sun,&nbsp;Xingtuan Yang,&nbsp;Shengyao Jiang","doi":"10.1016/j.pnucene.2025.105835","DOIUrl":"10.1016/j.pnucene.2025.105835","url":null,"abstract":"<div><div>This paper systematically investigates the flow characteristics, separation performance, and pressure drop behaviour of an axial cycloidal-profiled swirl vane separator under low-humidity conditions for small modular reactors. Through comprehensive experiments on a self-designed test platform, covering three drainage heights (300/400/500 mm) and five liquid flow rates (18–41.4 kg/h), the critical transition boundary from swirling annular flow to churn flow under low Reynolds numbers was revealed to exhibit linear growth characteristics, while existing theoretical models significantly overpredicted values in this operational range. In stable swirling annular flow regimes, the separation ratio (<em>η</em>) gradually decreases due to secondary droplet breakup. Flow regime transition triggers a sharp decline when liquid holdup (<em>ϕ</em>) exceeds the critical threshold (<em>ϕ</em>_<em>cri</em>). Increasing drainage height reduces <em>ϕ</em>_<em>cri</em> by approximately 10.2 % and narrows the stable operational window. Sobol global sensitivity analysis identifies liquid superficial velocity (<em>U</em>_<em>l</em>) as the dominant parameter influencing <em>η</em>, with dimensionless drainage height indirectly affecting <em>η</em> through coupling with gas velocity (<em>U</em>_<em>g</em>). Δ<em>P</em> exhibits a strong linear correlation with gas superficial velocity, and a localized “negative growth” phenomenon occurs at low <em>U</em>_<em>g</em> due to liquid film backflow. Non-gravitational pressure drop gradient ((d<em>P/</em>d<em>L</em>)_<em>f+a</em>) increases linearly with the gas-liquid velocity ratio, while a higher liquid flow rate elevates the pressure drop gradient. The proposed regression model validated the negative effect of <em>H∗</em> and the interaction effect of gas-liquid shear effects and liquid film inertia on (d<em>P/</em>d<em>L</em>)_<em>f+a</em>.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"187 ","pages":"Article 105835"},"PeriodicalIF":3.3,"publicationDate":"2025-05-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144138630","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":"Temperature gradient method for extracting background temperature and calculating temperature rise area of nuclear power plants' thermal discharge","authors":"Xinglong Guo ,&nbsp;Guoqing Li ,&nbsp;Luyan Wang ,&nbsp;Chao Sui ,&nbsp;Jun Zhu ,&nbsp;Hongyuan Shi ,&nbsp;Chao Zhan ,&nbsp;Qing Wang","doi":"10.1016/j.pnucene.2025.105850","DOIUrl":"10.1016/j.pnucene.2025.105850","url":null,"abstract":"<div><div>Remote sensing has become an essential tool for monitoring the discharge range and thermal discharge temperature rise classification of nuclear power plants. The core of remote sensing-based thermal discharge monitoring is the accurate extraction of background water temperature. Existing methods for extracting background temperature mainly involve two approaches: one is based on expert prior knowledge to define a temperature range, with the average temperature within this range used as the background temperature. However, this method is somewhat arbitrary and heavily influenced by human judgment. The second approach is based on deep learning, which can accurately extract background temperature but requires a large amount of training data and needs to be retrained for different datasets. To address these issues, we propose a background temperature extraction method based on the temperature gradient algorithm. To validate the applicability and accuracy of this method, we utilized 991 scenes of Landsat data from January 1, 2023, to June 30, 2024, and performed validation across 65 nuclear power plants worldwide. The results show that, compared to methods such as the average temperature correction and adjacent-zone substitution method, the temperature gradient method can automatically and accurately extract background temperature and temperature rise areas. Moreover, this method demonstrates strong general applicability, making it suitable for both coastal and lakeside nuclear power plants.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"187 ","pages":"Article 105850"},"PeriodicalIF":3.3,"publicationDate":"2025-05-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144134982","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":"Efficient calculation of reactor noise via Ito–Langevin Process for correlated fluctuations","authors":"Guy Gabrieli ,&nbsp;Yair Shokef ,&nbsp;Izhar Neder","doi":"10.1016/j.pnucene.2025.105822","DOIUrl":"10.1016/j.pnucene.2025.105822","url":null,"abstract":"<div><div>We derive an Ito–Langevin stochastic process that captures the time-dependent deviation from Poisson behavior of the noise detected from a general, spatially heterogeneous sub-critical neutron system. Using the probability generating function for the actual physical process, we deduce the super-Poisson deviation of the covariance matrix of counts at the detector due to neutron multiplication upon fission. This leads to a general form that coincides with the second moment of an Ito process. This comparison facilitates the formulation of a corresponding effective Langevin equation, which potentially enables simulations that significantly reduce the computational resources required compared to direct simulation of the system’s actual noise. This method could assist in designing sub-critical noise experiments for licensing new research reactors, for improving cross-section libraries and for non-destructive assays of spent fuel.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"187 ","pages":"Article 105822"},"PeriodicalIF":3.3,"publicationDate":"2025-05-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144107377","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}