Progress in Nuclear Energy最新文献

{"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}

{"title":"Safety margins improvement by means of the passive heat removal system and the HA-2 in VVER-1000/V320 reactors","authors":"Elena Redondo-Valero ,&nbsp;Cesar Queral ,&nbsp;Kevin Fernandez-Cosials ,&nbsp;Victor Hugo Sanchez-Espinoza","doi":"10.1016/j.pnucene.2025.105825","DOIUrl":"10.1016/j.pnucene.2025.105825","url":null,"abstract":"<div><div>One of the key functions of the safety systems is the removal of decay heat during an accidental sequence. Conventional VVER reactor designs have safety systems capable of removing residual heat in an emergency scenario, but these systems will be challenged to operate under station blackout conditions. As a result, some of the Gen III/Gen III + VVER designs incorporate passive safety systems capable of maintaining the decay heat removal in the event of a total loss of AC power. The present study focuses on the analysis of the air-cooled Passive Heat Removal System (PHRS) incorporated in some Gen III/Gen III + designs such as the VVER-1000/V412, the VVER-1200/V392M, V509, V523 or the VVER-TOI. For this purpose, a PHRS model has been developed in the TRACE system code, which has been incorporated into a model of a VVER-1000/V320 reactor which also includes another Gen III/Gen III + VVER reactors common feature; the Second Stage Hydroaccumulators (HA-2), subsystem of the Emergency Core Cooling System. Subsequently, a Station Blackout (SBO), a SBO along with an SBLOCA and a SBO along with an LBLOCA sequences have been analyzed with the air-cooled PHRS operating. The results show that in some scenarios, the PHRS performance is critical to lead the sequence to a safe state by transporting the decay heat to the atmosphere heat sink for at least 24 h.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"187 ","pages":"Article 105825"},"PeriodicalIF":3.3,"publicationDate":"2025-05-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143937804","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":"Research on the prediction of transient parameters in rod bundle channel based on POD-ML method","authors":"Yujie Xu ,&nbsp;Xiaomeng Dong ,&nbsp;Anqi Xu ,&nbsp;Jinhong Mo ,&nbsp;Yong Liu ,&nbsp;Ming Yang","doi":"10.1016/j.pnucene.2025.105832","DOIUrl":"10.1016/j.pnucene.2025.105832","url":null,"abstract":"<div><div>Reduced-order modeling (ROM) has been widely used to reduce the complexity of physical models by mapping full-order conservation equations to lower-order subspaces or constructing data-driven surrogate models. Compared with traditional computational fluid dynamics (CFD) simulations, reduced-order modeling is more computationally efficient in large-scale simulation calculations. In this study, we propose a reduced-order modeling framework under transient conditions using a combination of Proper Orthogonal Decomposition (POD) and machine learning (ML), which is used to implement the transient prediction of parameters in the channel of a rod bundle. The comparison results of the two different prediction methods show that the LSTM + POD method is more suitable for analyzing the short prediction of the simple varying temperature distribution and z-direction mass flow distribution. The prediction under complex conditions and the long prediction of the z-direction mass flow distribution are not as effective as the POD + LSTM method, which can provide a solution for the prediction of other transient systems in the future.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"187 ","pages":"Article 105832"},"PeriodicalIF":3.3,"publicationDate":"2025-05-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143941907","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":"Design and full core fuel performance assessment of high burnup cores for 4-loop PWRs","authors":"Assil Halimi ,&nbsp;Yifeng Che ,&nbsp;Koroush Shirvan","doi":"10.1016/j.pnucene.2025.105791","DOIUrl":"10.1016/j.pnucene.2025.105791","url":null,"abstract":"<div><div>Increasing the fuel discharge burnup of current light water reactors (LWRs) promises reductions in fuel cycle and/or operations costs. By assuming a constant core power density, the economic gain is enabled by better fuel utilization and/or an increased capacity factor. In this effort to investigate greater than 62 MWd/kgU maximum rod average burnup for 110+ kW/l core power density, two core designs have been developed for a standard 17x17, 193 fuel assemblies pressurized water reactor (PWR). The levelized unit cost methodology is employed to evaluate fuel cycle, operation and maintenance, and capital cost impacts and to examine the economic viability of both core design pathways. Core design and optimization are performed using the commercial STUDSVIK code package. Fuel performance analysis is realized in full core configuration via auditing FRAPCON4.1, FAST1.2, and the high-fidelity code BISON. To provide a realistic assessment, the core design process takes into consideration best practices in current PWR core design. It features acceptable performance in terms of various core design constraints on maximum allowable peaking and boron concentration. Gadolinia (Gd2O3) is used as a burnable poison with a maximum of 9 wt% concentration while feeding 89 or 77 fuel assemblies in a 3-batch refueling scheme. Full core fuel performance simulation, which allows for characterization of relevant fuel temperatures, plenum pressures, stresses, and strains, is performed with respect to two bounding burnup levels. Such performance is potentially licensable for the 18-month high burnup core (&lt;68 MWd/kgU peak pin), while it is more challenging for the 24-month high burnup core design pathway (&lt;75 MWd/kgU peak pin). Maximum rod plenum pressure is identified as the most limiting fuel performance parameter. While the scope of the present study focuses on the steady-state plus overpower conditions, the acceptability of the new discharge burnup has to be further assessed by considering uncertainties and impacts under accident scenarios in the future.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"186 ","pages":"Article 105791"},"PeriodicalIF":3.3,"publicationDate":"2025-05-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143931318","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":"Comprehensive analysis of oxygen mass transfer and impurity dynamics in oxygen-controlled ELSY lead-cooled reactors","authors":"Dingsheng Lu,&nbsp;Xiao Pang,&nbsp;Yupeng Yang,&nbsp;Chenglong Wang,&nbsp;Dalin Zhang,&nbsp;Suizheng Qiu","doi":"10.1016/j.pnucene.2025.105830","DOIUrl":"10.1016/j.pnucene.2025.105830","url":null,"abstract":"<div><div>In lead-cooled fast reactors, precise oxygen control is critical for mitigating corrosion, maintaining protective oxide layers, enhancing heat transfer, and reducing impurity deposition. However, achieving consistent oxygen management is challenging due to the high cost and limited availability of sensitive sensors, along with difficulties in monitoring oxygen distribution in pool-type reactors. Existing studies often oversimplify oxygen dynamics, neglecting the impacts of temperature gradients and cyclic flows on reactor-wide oxygen behavior and corrosion processes. This study refines traditional models by incorporating corrosion mechanisms and experimental data to analyze oxygen interactions with iron oxide nucleation in oxygen-controlled environments. Using Computational Fluid Dynamics (CFD), the distribution of oxygen and corrosion products is examined under various conditions. Results reveal hypoxic zones in high-flow and reflux regions under anoxic conditions, with inlet oxygen concentrations reaching a minimum at 1715 s. Enhanced gas-phase mass transfer stabilizes oxygen levels, reducing iron release and ensuring uniform distribution over time. Impurity nucleation begins in the cold zone beneath the steam generator at 7960 s, highlighting critical areas for management. By coupling reactor-wide oxygen dynamics and impurity behavior, this study underscores the importance of optimized oxygen management to mitigate corrosion, control impurities, and ensure long-term reactor stability, safety, and efficiency.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"186 ","pages":"Article 105830"},"PeriodicalIF":3.3,"publicationDate":"2025-05-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143937574","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":"Forecasting in-core power distributions in nuclear power plants via a spatial–temporal hierarchical-directed network","authors":"Weiqing Lin ,&nbsp;Xiren Miao ,&nbsp;Jing Chen ,&nbsp;Pengbin Duan ,&nbsp;Mingxin Ye ,&nbsp;Yong Xu ,&nbsp;Hao Jiang ,&nbsp;Yanzhen Lu","doi":"10.1016/j.pnucene.2025.105795","DOIUrl":"10.1016/j.pnucene.2025.105795","url":null,"abstract":"<div><div>As nuclear power plants (NPPs) undertake more peak regulation tasks to handle high new energy penetration and overcapacity, precise forecasting of in-core power distributions is essential for optimal control and safe operation. However, current works lack an effective strategy for predicting high-resolution power distributions and neglect in-core spatial correlations. This study proposes a spatial–temporal hierarchical-directed network (ST-HDN) for forecasting power distributions, whose prediction strategy is guided by the physical model. To characterize spatial correlations and causal relationships among physical quantities, the hierarchical-directed graph is designed and combined with neutron and power signals for input to the ST-HDN. Concretely, the ST-HDN integrates three sub-modules: a temporal-differencing layer to enhance representation of subtle variations; a multi-dilated convolutional network to extract dynamic temporal features; and a graph convolutional network to propagate spatial adjacent information, further predicting power nodes at various positions. The predicted power nodes are post-processed to derive future power distributions. Experiments on two peak regulation scenarios from a real-world NPP illustrate that the ST-HDN outperforms various state-of-the-art methods in 10-, 20-, and 30-min ahead forecasting.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"186 ","pages":"Article 105795"},"PeriodicalIF":3.3,"publicationDate":"2025-05-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143931841","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}