Nuclear Engineering and Design最新文献

{"title":"A review on PAR operating characteristics and impact on hydrogen behaviors in containment of PWRs","authors":"Tianming Man ,&nbsp;Youcai Feng ,&nbsp;Xuhua Zhou ,&nbsp;Zehua Guo ,&nbsp;Ming Ding","doi":"10.1016/j.nucengdes.2024.113806","DOIUrl":"10.1016/j.nucengdes.2024.113806","url":null,"abstract":"<div><div>Passive autocatalytic recombiners (PARs), owing to its excellent hydrogen removal capability and passive characteristic, have been widely utilized to prevent the potential hydrogen combustion risk in modern PWRs. The analysis of PAR operational characteristics has initially been conducted based on experimental investigations. Subsequently, it has gained the capability to comprehensively study the PAR operational behavior combing with the thermal–hydraulic characteristic and chemical reaction dynamics with the development in reaction kinetics models and advanced computing methods. It is significant to summarize the PAR development for improving the PAR operational behavior and mitigating the hydrogen combustion risk in the containment of PWRs. This paper summarized the operating principles and structural effects of PAR, the PAR operational behavior in various operating conditions, and the interaction between PARs and containment atmosphere in PWRs. Existing research provide substantial experimental and numerical simulation support for examining PAR operational behavior in the accident conditions. However, it remains necessary to conduct relevant researches to effectively use PAR for enhancing hydrogen risk mitigation capability in PWRs. Anticipated research includes improving PAR hydrogen removal capability, evaluating and enhancing PAR operational behavior in complex environments, and providing deeper insights for PARs impacts on the containment atmosphere in the future.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"432 ","pages":"Article 113806"},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143168317","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 and experimental study of shut-off rod assembly of a typical Indian research reactor under multi-support seismic excitation","authors":"A.Ravi Kiran ,&nbsp;M.K. Agrawal ,&nbsp;S.K. Sinha","doi":"10.1016/j.nucengdes.2025.113843","DOIUrl":"10.1016/j.nucengdes.2025.113843","url":null,"abstract":"<div><div>The Shut-off Rod (SoR) assembly is a critical safety-related system in the majority of nuclear reactors. For its seismic qualification, it is required to ensure that the Shut-off Rod falls freely through the guide tube in the stipulated time under seismic excitation. Due to this stringent requirement, full-scale testing is usually performed for seismic qualification. The present work demonstrates the methodology for the evaluation of functionality and integrity of SoR assembly under multi-support seismic excitation. Experimental and numerical studies are carried out on the SoR assembly of a typical Indian Research Reactor under multi-support seismic excitation. In the present work, the adequacy of numerical simulation for seismic qualification of the SoR assembly is explored. The present study provides two conditions to ensure the sufficiency of numerical simulation for the seismic qualification of the SoR assembly. The first condition is to ensure the structural integrity of the lattice guide tube and the second one is to ensure the free fall of the SoR. This paper provides a comprehensive overview of both experimental and numerical results, along with the sufficiency of numerical simulation for seismic qualification of the SoR assembly.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"432 ","pages":"Article 113843"},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143168320","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 soft measurement of the flow by reactor coolant pumps based on the motor current signature analysis","authors":"Zhaoliang Ding ,&nbsp;Xiuli Wang ,&nbsp;Yucan Zhang ,&nbsp;Yuanyuan Zhao ,&nbsp;Wei Xu","doi":"10.1016/j.nucengdes.2024.113797","DOIUrl":"10.1016/j.nucengdes.2024.113797","url":null,"abstract":"<div><div>Effective monitoring and identification of the operational flow of reactor coolant pumps (RCPs) are crucial for enhancing the safety and stability of nuclear power operations. To achieve the goal without the potential interference from intrusive sensors, the paper employs wavelet analysis and intrinsic time scale decomposition (ITD) to examine the collected current signals in both the time and frequency domains. The recurrent neural network (RNN) is used to determine the operational status of the pumps at different flow rates. The soft measurement model of the RCP flow, based on motor current signature analysis (MCSA), is built and experimentally validated using the RNN to identify the pump’s performance at different flow rates. The results show that the dominant frequency of pressure pulsation corresponds to the axial frequency in impeller channels, with the amplitude of the frequency component exhibiting a direct positive correlation to the flow rate. The impeller channel-generated pressure pulsation signal’s dominant frequency corresponds to the blade passing frequency, which exhibits an amplitude increase with higher flow rates. The soft measurement model, after the RNN training, achieves a recognition accuracy as high as 95.3% for flow rates of 0.8Q_d, 0.9Q_d, 1.0Q_d, 1.1Q_d, and 1.2Q_d. The model provides a certain value for subsequent studies in the field of pump flow monitoring.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"432 ","pages":"Article 113797"},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143168321","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":"The public radiological risk assessment for a generic HPR1000 in China","authors":"Mengxi Wang ,&nbsp;Hongtao Lin ,&nbsp;Liang Long ,&nbsp;Xinjian Liu ,&nbsp;Na Xue","doi":"10.1016/j.nucengdes.2024.113779","DOIUrl":"10.1016/j.nucengdes.2024.113779","url":null,"abstract":"<div><div>The public radiological risk assessment after severe accidents is considered a reasonable and comprehensive metric to determine the safety of a nuclear power plant, and it has attracted increasing attention since the Fukushima accident. Based on a generic HPR1000 in China, utilizing the source terms of Level-2 PSA and the site hourly meteorological data, individual doses and risks were simulated using a self-developed Level-3 PSA platform. The risk values were compared with the acceptability criteria of US NRC, UK and AR3.1.3. The results show that for both the most frequent and the most severe accident sequences, using the 90th percentile meteorological conditions, the doses are all below the dose limits specified in the Chinese national standard within a 5 km radius from the release point. For each release category, the exceedance frequency corresponding to the 7-day effective dose limit of 50 mSv within 3 km from the site is less than 30 %. The risk of each release category at the site boundary across the average meteorological conditions, and the annual average risks corresponding to the entire accident spectrum at different distances are all less than 10⁻⁸/reactor year, demonstrating good compliance with international risk acceptance criteria.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"432 ","pages":"Article 113779"},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143168323","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":"Comparative radiological impact of LOCA and RDD scenarios: An AI-enhanced assessment using HotSpot code","authors":"Najeeb N.M. Maglas ,&nbsp;Merouane Najar ,&nbsp;Zhao Qiang ,&nbsp;Mohsen M.M. Ali ,&nbsp;Ahmed AL-Osta ,&nbsp;M. Salah Alwarqi ,&nbsp;Djebara Lilia ,&nbsp;Alaa Fadul","doi":"10.1016/j.nucengdes.2024.113808","DOIUrl":"10.1016/j.nucengdes.2024.113808","url":null,"abstract":"<div><div>This study provides a comprehensive radiological assessment of two hypothetical incidents: a Loss of Coolant Accident (LOCA) at a nuclear reactor and a Radiological Dispersal Device (RDD) detonation, both simulated in Dhamar City, Yemen, using the HotSpot Health Physics Code. We evaluated the dispersion of radioactive materials under consistent atmospheric conditions to assess their environmental and human health impacts. Our analysis was based on two parameters: sampling time and exposure duration. For sampling time, the Total Effective Dose Equivalent (TEDE) was measured at specific intervals. After 2000 min, the TEDE for LOCA was 47 Sv, significantly higher than 0.0033 Sv for the RDD within a 1 km² area. In the initial moments of the explosion, the doses were 340 Sv for LOCA and 0.042 Sv for RDD, showing a dramatic decrease over time. For exposure duration, the LOCA scenario, results in a TEDE of 150 Sv after one year. In contrast, the RDD leads to a TEDE of 0.17 Sv after the same period. The LOCA scenario results in higher radiation doses due to multiple radionuclides with varying decay rates, causing a rapid increase in dose. In contrast, the RDD scenario shows a slower dose accumulation due to the long half-life of ¹³⁷Cs. This study introduces an AI-enhanced approach to radiological assessments of LOCA and RDD incidents, using an Artificial Neural Network (ANN) model comprised of classification and regression sub-models. The classification sub-model accurately identifies the nature of the radiation event, while the regression sub-model estimates the distance of the explosion within 80 km radius from the explosion epicenter. With a predictive accuracy of 100 % in classification and over 99 % in regression, the model significantly improves the effectiveness and speed of emergency response strategies, offering critical advancements in radiological safety measures. The impact on human organs was more severe in LOCA, with doses to the liver, skin, lungs, thyroid gland, brain, and kidneys exceeding those from the RDD by factors ranging from 55 to 6000. The findings stress the need for strong safety measures, long-term monitoring, and preparedness, especially in regions like Yemen, while highlighting the potential long-term environmental and health impacts of nuclear incidents and the importance of effective response and recovery plans.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"432 ","pages":"Article 113808"},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143168386","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":"Reactor power control performance improvement based on fuzzy neural network of a small pressurized water reactor","authors":"Juan Liu,&nbsp;Ge Gao,&nbsp;Ziqi Fan,&nbsp;Changhao Wu,&nbsp;Peiwei Sun,&nbsp;Xinyu Wei","doi":"10.1016/j.nucengdes.2024.113730","DOIUrl":"10.1016/j.nucengdes.2024.113730","url":null,"abstract":"<div><div>The modular design of small reactors has potential application prospects in nuclear power plants such as surface ships and land-based applications, and requires higher safety and flexibility. The ship nuclear power plant has the requirement of maneuvering and fast operation. As the power source, the small pressurized water reactor adopts the traditional control technology represented by PID (proportional-integral–differential), which cannot effectively meet the requirements of different operating conditions for control performance, especially the overshoot of reactor power is too large. Therefore, advanced control algorithms are needed to improve the control performance. In order to improve the learning and adaptive ability of fuzzy control, this paper takes a small PWR with 150 MW as the research object and proposes a reactor power control system based on fuzzy neural network. The simulation test results show that compared with fuzzy control and traditional control, the fuzzy neural network control has better learning and adaptive ability in the transient operation process, can adapt to various working conditions and effectively reduce the overshoot of reactor power. The overshoot of reactor power can be reduced by more than 15 % compared with the traditional control, and can be reduced by more than 8 % compared with the fuzzy control, and the control rod travel is also significantly reduced.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"432 ","pages":"Article 113730"},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143168388","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":"Preliminary thermo-mechanical coupling performance analysis of a three-lobe petal-shaped fuel rod","authors":"Binxian He ,&nbsp;Shusong Qin ,&nbsp;Aobo Han ,&nbsp;Wenchao Zhang ,&nbsp;Lipeng Du ,&nbsp;Xiangfei Meng ,&nbsp;Jianchuang Sun ,&nbsp;Weihua Cai","doi":"10.1016/j.nucengdes.2024.113792","DOIUrl":"10.1016/j.nucengdes.2024.113792","url":null,"abstract":"<div><div>In this paper, an analytical model of thermal–mechanical coupling performance of three-lobe and four-lobe petal-shaped fuel rods (PSFRs) irradiation is established based on the finite element method, and the thermo-mechanical characteristics of PSFR during normal operation and RIA condition are investigated. The results show that the temperature of three-lobe PSFR is large in the middle and small on both sides in the axial direction, and the petal lobe root is the largest in the circumferential direction. During normal operation, the cladding stress increases continuously with time, and the stress in the center region is the largest. At the 200th day and the 600th day, when RIA occurs suddenly, the cladding will exceed the yield strength in 1.003 s and 0.763 s, respectively. The cladding lobe root at the center of three-lobe PSFR is the location of maximum plastic deformation. There are similarities and differences in temperature, heat flux, and stress between three-lobe and four-lobe PSFRs, but they can be substituted for each other. The above study provides guidance for practical application of three-lobe PSFR.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"432 ","pages":"Article 113792"},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143169073","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":"The role of CFD combustion modelling in hydrogen safety management — IX: Validation of ETFC model implementation in flameFoam for large-scale hydrogen-air-steam deflagration","authors":"Julius Venckus ,&nbsp;Mantas Povilaitis","doi":"10.1016/j.nucengdes.2024.113737","DOIUrl":"10.1016/j.nucengdes.2024.113737","url":null,"abstract":"<div><div>In the context of CFD simulations for containment safety, Reynolds-Averaged Navier–Stokes (RANS) simulations are the most feasible given the large volume to be simulated. The combustion source term in these cases can be modelled by Turbulent Flame speed Closure (TFC) model, which is designed for fast combustion regimes and might fail to capture flame characteristics at slower regimes. This limitation of the TFC is addressed in an extended model version, the Extended Turbulent Flame speed Closure (ETFC). It should perform just as well in the fast regimes, but also be more accurate in the slower ones. Few studies have attempted to validate the extended model, and the existing ones only obtained good agreement with experiment when implementing user tuned weights. In this study we aim to validate the ETFC model against an experiment without modifying the equations.</div><div>CFD simulations of the Upward Propagating Flame Experiment in Hyka A2 experimental facility were performed using the combustion solver flameFoam with both the ETFC and TFC models for comparison. The obtained pressure evolution indicates faster flame propagation prediction by ETFC. Combustion was complete 0.6<!--> <!-->s later than in the experiment when using the ETFC, whereas the TFC was late by 0.9<!--> <!-->s. The peak <span><math><mrow><mi>d</mi><mi>p</mi><mo>/</mo><mi>d</mi><mi>t</mi></mrow></math></span> value was overestimated by 60<!--> <!-->% in the TFC case, but only by 37<!--> <!-->% with ETFC. A significant qualitative difference is a second <span><math><mrow><mi>d</mi><mi>p</mi><mo>/</mo><mi>d</mi><mi>t</mi></mrow></math></span> peak, exhibited by the TFC, whereas both the ETFC and the experiment only have one. Flame arrival time analysis reveals that both models reproduce the radial position well in a sense that flames are in the vicinity of thermocouples at experimentally measured times, although the exact positions are somewhat missed. The vertical position, however, was strongly underestimated by the TFC and captured well with the ETFC. Lastly, it was demonstrated that the ETFC combustion rate is sensitive to initial turbulence levels, with higher sensitivity to the turbulent kinetic energy.</div><div>Overall, this study demonstrates improved prediction capability of the ETFC in comparison to the TFC both for the flame shape and peak <span><math><mrow><mi>d</mi><mi>p</mi><mo>/</mo><mi>d</mi><mi>t</mi></mrow></math></span> values, and, unlike previous studies, good agreement with the experiment was achieved without tuning the model to match it.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"432 ","pages":"Article 113737"},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143167282","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":"An Artificial Neural Network (ANN) Model to Predict Critical Heat Flux (CHF) in a CANDU Fuel Element Simulation (FES) with Various Nonuniform Axial Heat Flux Shapes and Flow Liner Creep Profiles","authors":"F. Abbasian ,&nbsp;G.I. Hadaller ,&nbsp;R.A. Fortman ,&nbsp;J. Snell ,&nbsp;S. Park","doi":"10.1016/j.nucengdes.2024.113736","DOIUrl":"10.1016/j.nucengdes.2024.113736","url":null,"abstract":"<div><div>A classification ANN model was developed to predict critical power in a CANDU Fuel Element Simulation (FES) with various Axial Heat Flux Distributions (AFDs) and flow liner creep profiles. The ANN model employs 29 input features to model the AFDs and liner creep profiles and was trained by 433 test data. The classification ANN model was benchmarked against a standard regression ANN model developed with TensorFlow and the results are presented in this paper. The two models delivered roughly the same level of accuracy with a Root Mean Square Error (RMSE) of ∼ 2.5 %; however, the methodology used in the classification model seems to be able to alleviate overfitting and create a more tangible robustness in comparison with the regression model, albeit at the cost of a longer solution time. It is therefore recommended that the presented classification model be used in conjunction with typical regression models to attain more reliability, especially in problems including many features and small training datasets.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"432 ","pages":"Article 113736"},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143167514","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"The study of natural convection with decay heat source in an open horizontal neutron production target","authors":"Jianfei Tong ,&nbsp;Jiahui Chen ,&nbsp;Shenqiang Wei ,&nbsp;Songlin Wang ,&nbsp;Fei Shen ,&nbsp;Youlian Lu ,&nbsp;Ruiqiang Zhang ,&nbsp;Congju Yao ,&nbsp;Wen Yin ,&nbsp;Tianjiao Liang","doi":"10.1016/j.nucengdes.2024.113781","DOIUrl":"10.1016/j.nucengdes.2024.113781","url":null,"abstract":"<div><div>The research examines the phenomenon of natural convection heat transfer induced by decay heat sources within the intricate flow channels of the target. A three-dimensional model of the target was developed utilizing principles of fluid dynamics and numerical methods for heat transfer. The research emphasizes the micro-convective and non-flow heat transfer phenomena occurring within the confined spaces of the target’s intricate flow channels. Simulations of boundary conditions pertinent to a power failure incident were conducted and subsequently validated against experimental data. The findings indicated that the temperatures at the pressure inlet and pressure outlet boundary conditions closely aligned with the experimental measurements, recorded at 52.7 ℃ and 51.5 ℃, respectively, with an error margin of less than 3%. Additionally, the investigation revealed that the external surrounding heat transfer coefficient has a negligible effect on the internal temperature distribution. This research offers significant insights and recommendations for assessing accident scenarios at the target station for the China Spallation Neutron Source (CSNS) Ⅱ upgrade project.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"432 ","pages":"Article 113781"},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143167559","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}