Progress in Nuclear Energy最新文献

{"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}

{"title":"Numerical analysis of a helically coiled tube heat exchanger performance utilized in molten salt nuclear reactors for molten salt and He-CO2 working fluid conditions","authors":"Murat Öztürk ,&nbsp;Erdem Çiftçi","doi":"10.1016/j.pnucene.2025.105844","DOIUrl":"10.1016/j.pnucene.2025.105844","url":null,"abstract":"<div><div>In addition to the growing global energy demand, the intermittency of conventional renewable energy sources has increasingly directed attention toward nuclear energy. In this study, a heat exchanger responsible for circulating the coolant fluids in Molten Salt Reactors (MSR) was investigated, and the numerical analysis of a helically coiled tube heat exchanger was carried out. The numerical simulations were performed using the Computational Fluid Dynamics (CFD) approach and completed with ANSYS Fluent® software. Initially, analyses were conducted for pure CO₂ gas. Subsequently, CO₂ was mixed with 25 %, 50 %, and 75 % helium (He) gas, and analyses were completed for each case. Finally, pure He performance was evaluated. Each set of analyses was performed for Reynolds numbers (Re) of 50000 and 100000. The results indicate that an increase in the Re number significantly enhanced the heat transfer rate. The addition of helium notably increased the heat transfer rate, with the maximum rate being obtained using pure He gas. However, the transportation and cost of helium must also be considered. Ultimately, mixing CO₂ with He in the current model significantly improved system performance and could directly impact the power output of nuclear reactors. This approach enables higher coolant fluid temperatures to be achieved, thereby facilitating greater energy conversion efficiency.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"187 ","pages":"Article 105844"},"PeriodicalIF":3.3,"publicationDate":"2025-05-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144084602","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 the effect of drainage hooks on performance for wave-type vane separators of steam-water in PWR MSR","authors":"Qian Zhang ,&nbsp;Wenzhen Chen","doi":"10.1016/j.pnucene.2025.105824","DOIUrl":"10.1016/j.pnucene.2025.105824","url":null,"abstract":"<div><div>Extracting liquid water from the steam-water mixture by wave-type vanes is a key process in the moisture separator reheater (MSR) of pressurized water reactors (PWRs), which is essential to ensure turbine safety by reducing the water erosion damage and improve the relative internal efficiency of the turbine. Moreover, higher critical velocity of wave-type vanes, at which re-entrainment occurs, allows for smaller MSR equipment volume or a design with a significant margin for mass flow rate. Therefore, the moisture carryover, the pressure drop, and the critical velocity are the primary focus parameters in the development of wave-type vanes in MSR. In this work, two different drainage hook structures based on the same main wave-type vane were developed. Steam-water experiments with wave-type vanes in a dryer model under actual operating conditions for MSR were conducted to identify the drainage hook structure with the better separation characteristics. Experimental results on the moisture carryover, pressure drop, and the critical velocity were discussed to investigate the effect of different drainage hooks on separation characteristics. Under MSR operation conditions, the wave-type vane II exhibited lower pressure drop and higher separation efficiency across a broad range, making it a highly successful product recommended.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"187 ","pages":"Article 105824"},"PeriodicalIF":3.3,"publicationDate":"2025-05-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144084671","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 numerical study on single-phase heat transfer characteristics of the helical cruciform fuel assembly","authors":"Qi Zhang ,&nbsp;Yi Zhou ,&nbsp;Junsen Fu ,&nbsp;Hanyang Gu ,&nbsp;Tenglong Cong","doi":"10.1016/j.pnucene.2025.105826","DOIUrl":"10.1016/j.pnucene.2025.105826","url":null,"abstract":"<div><div>In the presented work, the single-phase heat transfer behaviors of the 4 × 4 helical cruciform fuel (HCF) assembly is investigated by numerical simulation. The non-uniform wall temperature of the HCF element and the evolution of mean wall temperature along the rod length are analyzed. Besides, the effects of helical pitch <span><math><mrow><msub><mi>L</mi><mrow><mi>H</mi><mi>P</mi></mrow></msub></mrow></math></span> and ratio of blade diameter and elbow diameter <span><math><mrow><msub><mi>D</mi><mi>R</mi></msub><mo>/</mo><msub><mi>D</mi><mi>r</mi></msub></mrow></math></span> on the heat transfer are paid particular concern. According to the results, the wall temperature continues to increase along the path between the elbow and the blade lee side, and the peak wall temperature is always observed at the end of blade lee side. While the blades of the neighboring HCF elements getting closer, the mean wall temperature is increased by 1.2 °C–4.4 °C rapidly. The hot spots are generated with the rotation angle of <em>n</em> × 90°. By increasing the rotation angle, the heat transfer coefficient (HTC) of the HCF assembly continues to fluctuate with the maximum amplitude of ±10 %, and the rotation angle for the fluctuation period is 90°. With shorter helical pitch, the mean wall temperature is reduced while more hot spots are generated. In addition, the distribution of mean wall temperature along the rod length is almost not changed by the cross-section shape.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"187 ","pages":"Article 105826"},"PeriodicalIF":3.3,"publicationDate":"2025-05-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144071398","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":"Investigation on effect of mixing factor on developing CHF correlation for fuel assembly","authors":"Shuo Chen ,&nbsp;Wei Zhang ,&nbsp;Junsen Fu ,&nbsp;Chenwei Liu ,&nbsp;Yao Xiao ,&nbsp;Hanyang Gu","doi":"10.1016/j.pnucene.2025.105836","DOIUrl":"10.1016/j.pnucene.2025.105836","url":null,"abstract":"<div><div>The Safety Margin in CHF correlation development is one important criterion of reactor safety analysis. Current critical heat flux correlation development relies on the accuracy of input thermal-hydraulic parameters, leading to the difficulty of reducing uncertainty. The turbulent mixing, which is simply modeled as a diffusion process, has a great influence on the distribution of local parameters and the safety margin of a certain correlation. To find the influence of the subchannel analysis program on CHF correlation development, In this paper, a critical heat flux (CHF) correlation is developed using the minimum departure of nucleate boiling ratio (MDNBR) method based on the subchannel code. The DNBR limit is determined using the Owen criterion and statistically evaluated. Then analysis is conducted on the estimation performance, DNBR limits and actual maximum heat flux of the corresponding correlation for different mixing factors. Results shows that the mixing factor contributes to the redistribution of local parameters. Besides, the relationship is weak between the accuracy of the correlation prediction and the value of the mixing factor bias. However, CHF correlations developed based on precise mixing factors obtained from experiments have the best accuracy and safety margin. The work is meaningful for the refinement of mixing effects for grid spacers in subchannel codes.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"187 ","pages":"Article 105836"},"PeriodicalIF":3.3,"publicationDate":"2025-05-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144069166","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":"Investigation of three-dimensional effects on natural circulation nuclear reactors: Chimney recirculation","authors":"Vincenzo Zingales ,&nbsp;Francesco D'Auria ,&nbsp;Yassin A. Hassan ,&nbsp;Dominique Bestion ,&nbsp;Klaus Umminger","doi":"10.1016/j.pnucene.2025.105833","DOIUrl":"10.1016/j.pnucene.2025.105833","url":null,"abstract":"<div><div>This paper explores the intricate interplay of three-dimensional effects on the operational dynamics of nuclear reactors employing natural circulation (NC). A review of thermal-hydraulic phenomena (THP), with insights into three-dimensional THP and NC, as well as of their role in deterministic safety assessment has been performed. Moreover, some ongoing research directions relevant to NC reactors have been summarized, showing that they do not address the concerns highlighted in the present work. In this work, attention is directed towards reactors featuring a long chimney, also called riser, in which upward movement of hot fluid happens, together with a downflow with lower temperature in the surrounding downcomer. The study delves into the potential occurrence of buoyancy driven recirculation phenomena within the chimney and the consequential risk of cold fluid ingress into the core region. Such occurrences may precipitate in instabilities including the neutronic-thermal hydraulic feedback. In this work, we consider single phase flow including the presence of subcooled void and two-phase flow in the core region which are typical respectively of NC iPWRs and NC BWRs. Conceptually simplified RELAP5-3D nodalizations have been adopted, in which the chimney region has been divided into a peripheral annular part and a central cylindrical region. Recirculation in the chimney is affected by core power, which also determines NC flow inside the vessel. The performed study is preliminary considering that no experimental data is available, and the system code RELAP5-3D is used instead of a more powerful CFD code, perhaps more suited in case of single-phase conditions.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"187 ","pages":"Article 105833"},"PeriodicalIF":3.3,"publicationDate":"2025-05-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144069167","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}