Progress in Nuclear Energy最新文献

{"title":"Analysis of the MOTEL helical coil steam generator tests MS-SG01 and MS-SG02 using TRACE","authors":"Yago Martinez-Gonzalez ,&nbsp;Cesar Queral ,&nbsp;Jorge Sanchez-Torrijos ,&nbsp;Elena Redondo-Valero","doi":"10.1016/j.pnucene.2025.106014","DOIUrl":"10.1016/j.pnucene.2025.106014","url":null,"abstract":"<div><div>Small modular reactors (SMRs) are gaining relevance in the nuclear industry. In these reactors, the steam generators play a crucial role due to its more compact design compared to the steam generators of large power plants. The steam generators of an integral SMR are generally composed of multiple tubes arranged in a helical coil configuration. Understanding the thermal-hydraulic behavior of this type of steam generator is essential to ensure safe and efficient operation of SMRs.</div><div>The objective of this study is to validate the TRACE system code by comparing its results with the experimental data obtained at the MOTEL (MOdular Test Loop) facility, located in LUT (Finland), with a reactor coolant system configuration similar to the NuScale reactor design. The work focuses on the study of helical coil steam generators, deepening the understanding of their behavior in both stationary and transient conditions. Experimental tests MS-SG01 and MS-SG02 have been carried out in both medium/high power (250 kW, 500 kW, 750 kW and 1000 kW) and low power (75 kW, 100 kW, 125 kW and 150 kW) conditions. The simulation results show that the TRACE code accurately reproduces the main parameters of interest recorded during the experimental tests.</div><div>The simulation results show that the TRACE code accurately reproduces the main parameters of interest during the MOTEL tests. In particular, TRACE reproduces core inlet/outlet temperatures with relative errors below 0.32 % and primary mass flow rates within 7 % of the experimental data, showing its adequacy for SMR with helical steam generators.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"191 ","pages":"Article 106014"},"PeriodicalIF":3.2,"publicationDate":"2025-09-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144997490","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Development and economic analysis of fast breeder reactor- pressurized water reactor's two-component nuclear power system","authors":"Zhiwen Dai,&nbsp;Donghui Zhang,&nbsp;Fuhui Zhang,&nbsp;Shangang Cao,&nbsp;Songping Wang,&nbsp;Yuting Yang,&nbsp;Jintao Dong","doi":"10.1016/j.pnucene.2025.106011","DOIUrl":"10.1016/j.pnucene.2025.106011","url":null,"abstract":"<div><div>The global installed capacity of nuclear power is expected to grow significantly. Fast Breeder Reactors (FBRs) could offer an efficient, safe, and sustainable energy solution, supporting long-term nuclear power development while reducing nuclear waste. This paper develops scale estimation and economic analysis models for the two-component nuclear power system consisting of Pressurized Water Reactors (PWRs) and FBRs, using 1.2 GWe commercial fast reactors CFR1000 and BN-1200 as case studies. The development scales of PWRs and FBRs are estimated through case studies in China, considering single-reactor systems, once-through cycles, and two-component nuclear power systems. The results indicate that China's two-component nuclear power system could achieve a total installed capacity of 500 GWe. Economic modeling reveals Levelized Unit Energy Costs (LUEC) of 0.0459 USD/kWh for CFR1000 and 0.0494 USD/kWh for BN-1200, with a detailed sensitivity analysis of key impact factors. As the two-component nuclear power system continues to develop globally, fission power of PWR and FBR could serve as an affordable and sustainable baseload power in the future.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"190 ","pages":"Article 106011"},"PeriodicalIF":3.2,"publicationDate":"2025-09-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144932081","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":"Heat transfer enhancement using different Ag/GPE nanofluid mixing conditions on a downward-facing heating surface","authors":"Qi Wang,&nbsp;Huai-En Hsieh,&nbsp;Xintian Cai,&nbsp;Saikun Wang,&nbsp;Kaiwen Li,&nbsp;Chenrui Lin","doi":"10.1016/j.pnucene.2025.106012","DOIUrl":"10.1016/j.pnucene.2025.106012","url":null,"abstract":"<div><div>In nuclear engineering, effective heat dissipation is paramount for safety, making the enhancement of critical heat flux (CHF) using modified nanofluids a crucial area of research. While various nanoparticle combinations have been investigated, the impact of silver (Ag) and graphene (GPE) in downward-facing boiling remains largely unexplored. This study assessed five mass ratios (ranging from 4:4 to 12:12 mg/L) under controlled heating conditions, utilizing surface characterization and high-speed bubble tracking methodologies. The 8:8 ratio yielded the highest CHF, surpassing that of deionized water by 23.1 %, whereas the 12:12 ratio resulted in a 5.9 % decrease in CHF. The 12:12 configuration, characterized by a roughness of 0.294 μm, experienced challenges with vapor film stability attributed to particle aggregation. In contrast, the 8:8 configuration demonstrated a smoother and more uniform dispersion (roughness of 0.168 μm, contact angle of 59.334°). These observations highlight three interconnected mechanisms—enhanced wettability, improved thermal pathways, and bubble dynamics—that offer valuable insights for the design of hybrid nanofluids in gravity-assisted boiling systems. The substantial CHF enhancement at the optimal ratio advocates for the adoption of in-vessel retention (IVR) strategies, thereby bolstering heat dissipation capabilities during severe accident of nuclear reactors and reducing the likelihood of pressure vessel failure.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"190 ","pages":"Article 106012"},"PeriodicalIF":3.2,"publicationDate":"2025-09-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144932080","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":"VITARS platform: Development and application in nuclear power plant safety analysis","authors":"Qianxi Xiao ,&nbsp;Chao Wang ,&nbsp;Ronghua Chen ,&nbsp;Guoqing Lu ,&nbsp;Dezhi Wan ,&nbsp;Liqun Zhang ,&nbsp;Changjiang Yang ,&nbsp;Wenxi Tian ,&nbsp;Suizheng Qiu","doi":"10.1016/j.pnucene.2025.106013","DOIUrl":"10.1016/j.pnucene.2025.106013","url":null,"abstract":"<div><div>A Visual Interactive Transient Analysis code for nuclear Reactor System (VITARS) platform was developed to enhance the efficiency of nuclear reactor simulation analysis and replicate real-time operational conditions of reactor circuits. The comprehensive platform encompasses the entire simulation process, including model establishment, real-time operation, and post-processing analysis. It supports multi-process parallel computing and incorporates a specialized interface for data exchange with Distributed Control System (DCS) virtual machines. The platform enables graphical construction of thermal-hydraulic models, logic configuration, and operational interfaces. During simulation, interaction with DCS virtual machines facilitates the simulation of measurement points and equipment status, enabling the application and verification of DCS closed-loop control logic. To validate its functionality, simulation analyses were conducted with a three-loop comprehensive thermal-hydraulic test facility from China as the target. In this case study, all platform modules were effectively utilized. Steady-state results demonstrated that key model parameters could be stabilized at correct values through DCS Proportional-Integral-Derivative (PID) closed-loop logic control. Transient analysis verified the effectiveness of protection logic triggers and the accuracy of the thermal-hydraulic model. Results indicated that safety systems activated promptly, and parameter variations followed expected patterns, yielding reasonable outcomes. These findings collectively validate the platform's effectiveness in both modeling and operational phases.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"190 ","pages":"Article 106013"},"PeriodicalIF":3.2,"publicationDate":"2025-09-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144988618","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":"Proposal for a PWR-type vSMR for hydrogen-electricity-water desalination polygeneration","authors":"Daniel González Rodríguez ,&nbsp;Maritza Rodriguez Gual ,&nbsp;Leorlen Yunier Rojas Mazaira ,&nbsp;Antonio Celso Dantas Antonino ,&nbsp;Nathalia Araújo ,&nbsp;Marcos Maturana ,&nbsp;Carlos Alberto Brayner de Oliveira Lira","doi":"10.1016/j.pnucene.2025.105999","DOIUrl":"10.1016/j.pnucene.2025.105999","url":null,"abstract":"<div><div>In this study, a new polygeneration system of a Pressurized Water Reactor (PWR)-type vSMR with Multi-Stage Flash (MSF) desalination and Alkaline Water Electrolysis (AWE or AEC) for hydrogen production is considered, aiming to address water scarcity and contribute to energy source diversification and a sustainable future in Brazil. Three cogeneration systems were analyzed: PWR + MSF model for desalination, PWR + AEC model for hydrogen production, and a polygeneration model (PWR + AEC + MSF) combining both processes. Thermodynamic and exergetic analyses were conducted using Aspen Plus software to evaluate energy efficiency, exergy destruction, and system performance under various operating conditions. The PWR + MSF configuration demonstrated an energy efficiency of 23.12 % and an exergy efficiency of 56.18 %, producing 9.65 kg/s of desalinated water. The PWR + AEC system achieved a hydrogen production rate of 0.0923 kg/s, with an energy efficiency of 26.83 % and an exergy efficiency of 47.99 %. The polygeneration model (PWR + AEC + MSF) showed slightly reduced efficiencies due to increased system complexity but successfully produced both desalinated water (9.65 kg/s) and hydrogen (0.0923 kg/s). Overall efficiency results indicate that all configurations are thermodynamically viable, with the PWR + MSF system being the most efficient (31.38 %). The exergetic sustainability index of the polygeneration system is found to be 1.73. The sensitivity analysis of three proposed cogeneration plants is performed to determine the most efficient operating conditions. The performance of three cogeneration systems is evaluated thermodynamically. The study highlights the potential of microreactors for producing fresh water and hydrogen simultaneously in a sustainable way. The main findings, advantages, and comparisons of the proposed cogeneration systems and recommendations for future research directions are summarized. This work contributes to the growing research on nuclear polygeneration, offering insights into integrating microreactors with desalination and hydrogen production for a sustainable energy future.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"190 ","pages":"Article 105999"},"PeriodicalIF":3.2,"publicationDate":"2025-09-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144932079","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":"Assessing manufacturing defects in ceramic composites with both simulated and experimental synchrotron computed tomography","authors":"Iwan T. Mitchell ,&nbsp;Fabio Martini ,&nbsp;Gareth F. Stephens ,&nbsp;Simon C. Middleburgh ,&nbsp;Franck P. Vidal","doi":"10.1016/j.pnucene.2025.106002","DOIUrl":"10.1016/j.pnucene.2025.106002","url":null,"abstract":"<div><div>Non-destructive testing using X-ray computer tomography (XCT) has been used to assess the applicability of visualising ceramic kernels held within a dissimilar ceramic matrix. Simulations were performed to ascertain the feasibility of CT scans of such samples, and optimise the scanning parameters offline. Corresponding experiments were carried out to assess the defects in the structure that exist as a result of manufacturing methods in zirconium diboride (ZrB₂) kernels held within a zirconium dioxide (ZrO₂) matrix material. Ceramic–ceramic matrix composites are garnering a great deal of interest in a number of applications, including as nuclear fuels for high temperature gas reactors and the methodology presented has potential to be of use in assessing the position and state of pellets incorporated into a ceramic matrix, whilst being able to detect features such as cracks, porosity and interfaces between kernels and the matrix. Computer modelling of the composites supports the experimental observations and has been used to assess the plausibility of assessing a higher density of kernels held within a ceramic matrix that will support ongoing work, whilst highlighting a valid method for periodical assessment of fuel manufacturing processes.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"190 ","pages":"Article 106002"},"PeriodicalIF":3.2,"publicationDate":"2025-09-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144922939","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":"From black box to glass box: Explainable AI for enhancing operator decision making in reactor accident scenarios","authors":"Merouane Najar,&nbsp;He Wang","doi":"10.1016/j.pnucene.2025.106009","DOIUrl":"10.1016/j.pnucene.2025.106009","url":null,"abstract":"<div><div>The main objective of integrating Artificial Intelligence (AI) with nuclear safety systems is to enhance operator reliability during reactor accidents by optimizing their performance. AI is employed to provide operators with critical information, enabling swift, informed decisions that minimize errors and ensure the safe operation of Nuclear Power Plants (NPPs). However, real-world AI applications in nuclear engineering must first establish operator trust in the model's predictions. Operators need to understand the model's decision-making process, allowing them to use their knowledge and experience to validate its results.</div><div>To address this, the study employs sensor datasets from four reactor accidents, covering failure ranges between 1 % and 100 %. Three deep learning (DL) models with different capabilities are trained: an Artificial Neural Network (ANN) and Convolutional Neural Network (CNN) are used to classify accidents based on both numerical and image data, providing operators with detailed information about the event's nature. Additionally, Long Short-Term Memory (LSTM) networks are used to forecast the event's progression, enabling operators to assess future severity.</div><div>Furthermore, Explainable AI (XAI) techniques are applied to ensure transparency in the model prediction process, simplifying model complexity and transforming them from \"black boxes\" into \"glass boxes\", which allows operators to understand the rationale behind AI-driven decisions, thereby establishing trust in their outputs. The results show that both classification models demonstrate strong performance, with CNN achieving the highest accuracy at 97 %. Additionally, the LSTM model exhibited an R² score exceeding 96 %. The application of XAI, combined with innovative data visualization techniques, effectively explains the reasoning behind the models' predictions.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"190 ","pages":"Article 106009"},"PeriodicalIF":3.2,"publicationDate":"2025-08-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144920376","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":"Determination of scanning rate correction factor on the dose rate measured by a dosimeter in motion","authors":"Hyeonmin Lee ,&nbsp;Seung Hun Shin ,&nbsp;Hee Reyoung Kim","doi":"10.1016/j.pnucene.2025.106005","DOIUrl":"10.1016/j.pnucene.2025.106005","url":null,"abstract":"<div><div>This study analyzed the scanning rate correction of the dose rate using a mobile dosimeter. An accident at a nuclear power plant exposes nearby residents to external radiation from radioactive nuclides released into the atmosphere. Residents resort to indoor sheltering or evacuate using vehicles according to the disaster evacuation plan. Considering this process, a vehicle-mounted mobile dosimeter was developed to manage the radiation exposure of residents and monitor the dose along the evacuation route. Monitoring the on-site dose along the evacuation route using a mobile radiation detector mounted in a vehicle can cause dose rate underestimation depending on the scan speed. Compensating for the underestimated dose is necessary to determine the risk area from the dose measurement results and ensure greater safety in the process after the secondary evacuation. In this study, MCNP (Monte Carlo N-Particle Transport) 6.2 simulation based on experimental verification using a moving point source was employed to evaluate the degree of dose reduction according to the scan speed, and a function to compensate for the dose reduction was determined.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"190 ","pages":"Article 106005"},"PeriodicalIF":3.2,"publicationDate":"2025-08-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144916313","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":"Multidimensional ECC behaviors of air-water flow in a downcomer annulus with direct vessel injection mode during the LBLOCA reflood phase","authors":"Xin Ma ,&nbsp;Wan Sun ,&nbsp;Jie Wan ,&nbsp;Yang Liu ,&nbsp;Zhanghao Zhu ,&nbsp;Huafa Chen ,&nbsp;Luteng Zhang ,&nbsp;Zaiyong Ma ,&nbsp;Liangming Pan","doi":"10.1016/j.pnucene.2025.106003","DOIUrl":"10.1016/j.pnucene.2025.106003","url":null,"abstract":"<div><div>The direct vessel injection (DVI) technology has been increasingly adopted in new pressurized water reactors due to its advantages in simplifying the safety injection system design and enhancing economic benefits. This study focuses on the modified DVI safety injection system of the HPR1000. An improved linear scaling method was employed to model the prototype, and relevant experiments were conducted on a 1:8.5 scale visualization test section. During the refilling and reflooding stages of a large break loss of coolant accident (LBLOCA), phenomena such as counter current flow limitation (CCFL) and bypass flow near the break were observed within the annular cavity. The research investigated the impacts of various break locations and the presence or absence of guiding structures on the bypass effect in the DVI safety injection system. Additionally, comparisons were made between DVI safety injection and cold - leg injection. The findings reveal that direct bypass flow dominates the safety injection bypass during the refilling and reflooding stages of an LBLOCA. The closer the cold - leg break location is to the DVI nozzle, the more significant the increase in bypass flow at the break location. The installation of guiding devices can effectively reduce the proportion of safety injection bypass flow. The data results of this study provide crucial insights for the optimization and innovation of the modified safety injection system in HPR1000.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"190 ","pages":"Article 106003"},"PeriodicalIF":3.2,"publicationDate":"2025-08-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144908656","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":"Assessment of the lifecycle cost of nuclear-grade coolants for advanced reactors","authors":"Lucas Javier Fernandez de Losada ,&nbsp;Jacopo Buongiorno ,&nbsp;Annalisa Manera","doi":"10.1016/j.pnucene.2025.105888","DOIUrl":"10.1016/j.pnucene.2025.105888","url":null,"abstract":"<div><div>The vast majority of advanced nuclear energy systems (both fission and fusion) use coolant fluids other than water. While the thermophysical properties of all these coolants are well known, their cost, when the appropriate nuclear-grade characteristics are accounted for, is not. The fluids analyzed in this research are: light water (taken as reference), liquid sodium, liquid lead and lead-bismuth eutectic (LBE), organic coolants (terphenyl derivatives), liquid salt (FLiBe) and helium gas. A credible estimate of all the costs associated with such coolants must include raw material supply, transportation, initial purification to nuclear-grade specifications, chemistry control during operation and disposal. Here we analyze the current purchase cost of various types of reactor coolants considering the target purity required for the start of the nuclear operations in a nuclear power plant, including the cost of pre-treatment if necessary. We also perform an analysis of the costs required to control the coolant chemistry and purity throughout the reactor lifetime. To that end, all major expected impurities and the corresponding concentration limits have been identified for each coolant. Consequently, the state-of-the-art purification methods and related systems have been analyzed and their investment and operating costs have been estimated. The cost of coolant makeup has also been considered in the analysis where appropriate. The cost of disposal of the coolants at the backend of the reactor service life has been excluded due to its negligible discounted value. The final results of the research include the estimate of a CAPEX cost expressed in $/kg of coolant and an OPEX cost in $/year per megawatt.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"190 ","pages":"Article 105888"},"PeriodicalIF":3.2,"publicationDate":"2025-08-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144908549","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}